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| {{#Wiki_filter:.- ENERGY DUKE Brunswick Vice President William R. Gideon Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 December 21, 2015 Serial: BSEP 15-01 01 TSC-201 5-03 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 | | {{#Wiki_filter:}} |
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| ==Subject:==
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| Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Application For Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program Ladies and Gentlemen:
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| In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR Part 50.90), "Application for Amendment of License, Construction Permit, or Early Site Permit,"
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| Duke Energy Progress, Inc. (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BS EP), Unit Nos. 1 and 2. The proposed amendment would modify TS 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 Sb, Risk-Informed Method for Control of Surveillance Frequencies." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.5, "Programs and Manuals." The changes are consistent with Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) Standard Technical Specifications (STS) Change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative Sb," Revision 3 (ADAMS Accession No. ML090850642).
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| Enclosure 1 provides a description of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Enclosure 2 provides documentation of Probabilistic Risk Assessment (PRA) technical adequacy. Enclosure 3 and 4 provide the existing TS pages marked up to show the proposed changes for Units 1 and 2, respectively.
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| Enclosure 5 provides, for information only, the existing Unit 1 TS Bases pages marked up to show the proposed changes. Enclosure 6 provides a cross-reference between the TSTF-425 marked up TS pages and the BSEP Unit 1 and 2 TS pages. Enclosure 7 provides the No Significant Hazards Consideration.
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| Duke Energy requests approval of the proposed license amendment by December 21, 2016, with the amendment being implemented within 120 days.
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| In accordance with 10 CFR 50.91* Duke Energy is providing a copy of the proposed license amendment to the designated representative for the State of North Carolina.
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| U.S. Nuclear Regulatory Commission Page 2 of 3 Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.
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| I declare, under penalty of perjury, that the foregoing is true and correct. Executed on December 21, 2015.
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| William R. Gideon MAT/mat
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| ==Enclosures:==
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| : 1. Evaluation of Proposed License Amendment Request
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| : 2. Documentation of Probabilistic Risk Assessment (PRA) Technical Adequacy
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| : 3. Marked-up Technical Specification and Operating License Pages - Unit 1
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| : 4. Marked-up Technical Specification and Operating License Pages - Unit 2
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| : 5. Marked-up Technical Specification Bases Pages - Unit 1 (For Information Only)
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| : 6. Cross-Reference between TSTF-425, Revision 3 and BSEP Technical Specifications
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| : 7. Proposed No Significant Hazards Consideration Determination
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| U.S. Nuclear Regulatory Commission Page 3 of 3 cc (with enclosures):
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| U.S. Nuclear Regulatory Commission, Region II ATTN: Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWEN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)
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| Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1.645 lee.cox@ dhhs.nc.gov
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| BSEP 15-01 01 Enclosure 1 Page 1 of 4 Evaluation of Proposed License Amendment Request
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| ==Subject:==
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| Application For Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program 1.0 Description The proposed amendments would modify the Brunswick Steam Electric Plant (BSEP), Units Nos. 1 and 2, 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 5." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.5, "Programs and Manuals."
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| The changes are consistent with Nuclear Regulatory Commission (NRC) approved Industry/TISTF Standard Technical Specifications (STS) change TSTF-425, Revision 3 (ADAMS Accession No. ML090850642). The availability of the TS improvement was published in the FederalRegister on July 6, 2009 (74 FR 31996).
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| 2.0 Assessment 2.1 Applicability of Published Safety Evaluation Duke Energy Progress, Inc., (Duke Energy) has reviewed the safety evaluation dated July 6, 2009 (74 FR 31996). This review included a review of the NRC staff's evaluation, TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Revision 1 (ADAMS Accession No. ML071360456). includes Duke Energy's documentation with regard to Probabilistic Risk Assessment (PRA) technical adequacy consistent with the requirements of Section 4.2 of Regulatory Guide 1.200, Revision 2 (ADAMS Accession No. ML090410014), and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.
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| Duke Energy has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to BSEP and justify this amendment to incorporate the changes to the BSEP TSs.
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| 2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3, but Duke Energy proposes variations or deviations from TSTF-425, as identified below.
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| *Revised (i.e., typed) TS pages are not included in this amendment request given the number of TS pages affected and the straightforward nature of the proposed changes.
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| Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90, "Application for Amendment of License, Construction Permit, or Early
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| BSEP 15-01 01 Enclosure 1 Page 2 of 4 Site Permit," in that the mark-ups fully describe the changes desired. This represents an administrative deviation from the NRC staff's model application dated July 6, 2009 (74 FR 31996), with no impact on the NRC staff's model safety evaluation published in the same Federal Register Notice. As a result of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the NRC staff's model application. This deviation is consistent with many other industry applications adopting TSTF-425 (e.g., NRC Accession Nos.
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| ML14105A042 (i.e., Turkey Point) and ML14259A564 (i.e., Fermi)).
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| *The definition of STAGGERED TEST BASIS is being retained in BSEP TS Definition Section 1.1 because this terminology is used in TS Programs and Manuals Section 5.5.13, "Control Room Envelope Habitability Program," which is not the subject of this amendment request and is not proposed to be changed. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| *NRC letter dated April 14, 2010 (i.e., ML100990099), provides a change to an optional insert (Insert #2) to the existing TS Bases to facilitate adoption of the Traveler. The TSTF 425 TS Bases insert states the following:
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| The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
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| This statement only applies to frequencies that have been changed in accordance with the Surveillance Frequency Control Program (SFCP) and does not apply to frequencies that are relocated but not changed. Consistent with NUREG-1433, Revision 4 (i.e.,
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| ML12104A193), Duke Energy has replaced the TSTF-425 TS Bases Insert #2 with the following:
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| For Bases sections addressing multiple affected SRs, the TSTF-425 TS Bases Insert #2 has been revised to state the following:
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| The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
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| *Enclosure 6 provides a cross-reference between the NUREG-1433 Surveillance Requirements (SRs) included in TSTF-425 versus BSEP TS. This Enclosure includes a summary description of the referenced TSTF-425/BSEP TS SRs which is being provided for information purposes only and is not intended to be a verbatim description of the TS SRs. This cross-reference highlights the following:
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| o SRs included in TSTF-425 and corresponding BSEP SRs with identical SR numbers; o SRs included in TSTF-425 and corresponding BSEP SRs with differing SR numbers; o SRs included in TSTF-425 that are not contained in the BSEP TS; and o BSEP plant-specific SRs that are not contained in the TSTF- 425 mark-ups
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| BSEP 15-0101 Enclosure 1 Page 3 of 4 Concerning the above, BSEP SRs that have SR numbers identical to the corresponding TSTF-425 SRs are not deviations from TSTF-425. BSEP SRs with SR numbers that differ from the corresponding TSTF-425 SRs are administrative deviations from TSTF 425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| For TSTF-425 SRs that are not contained in the BSEP TS, the corresponding mark-ups included in TSTF-425 for these SRs are not applicable to BSEP. This is an administrative deviation from TSTF-425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| For BSEP plant-specific SRs that are not contained in the mark-ups provided in TSTF-425, Duke Energy has determined that the relocation of the frequencies for these BSEP plant specific SRs is consistent with the intent of TSTF-425, Revision 3, and with the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, "Introduction," of the model safety evaluation.
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| The subject plant-specific SRs involve fixed periodic frequencies. In accordance with TSTF-425, changes to the frequencies for these SRs would be controlled under the Surveillance Frequency Control Program. The Surveillance Frequency Control Program provides the necessary administrative controls to require that SRs related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Changes to frequencies in the Surveillance Frequency Control Program would be evaluated using the NRC approved methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1.
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| *SR 3.1.2.1 is included within the scope of this submittal but was not included in TSTF-425, Revision 3. The frequency of SR 3.1.2.1 is encompassed by the intent of TSTF-425 and, therefore, is within the scope of the NRC Model Safety Evaluation (i.e.,
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| ML091800157). The NUREG-1 433 markups within TSTF-425 include a similar core exposure based SR frequency (i.e., SR 3.3.1.1.6). During the NRC review of TSTF-425, Revision 1, an Request for Additional Information (RAI) response (i.e., ML080280272) from the TSTF specifically identified frequencies based on core exposure to be within the scope of TSTF-425 and NEI 04-10. In addition, on July 14, 2015, the NRC approved a similar SR frequency relocation for the Fermi TSTF-425 License Amendment (i.e.,
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| ML15155B416).
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| 3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Determination Duke Energy has reviewed the proposed no significant hazards consideration (NSHC) determination published in the Federal Register on July 6, 2009 (74 FR 31996) and has concluded that the proposed NSHC presented in the Federal Register notice is applicable to BSEP. As such, the NSHC determination for this amendment request is provided as to this license amendment request which satisfies the requirements of 10 CFR 50.91 (a).
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| BSEP 15-0101 Enclosure 1 Page 4 of 4 3.2 Commitments There are no new regulatory commitments contained in this submittal.
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| 3.3 Applicable Regulatory Requirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3, and the NRC's model safety evaluation published in the Notice of Availability dated July 6, 2009 (74 FR 31996). Duke Energy has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the Federal Register notice is applicable to BSEP.
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| 3.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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| 4.0 Environmental Consideration Duke Energy has reviewed the environmental consideration included in the NRC staff's model safety evaluation published in the Federal Register on July 6, 2009 (74 FR 31996). Duke Energy has concluded that the staff's findings presented therein are applicable to BSEP, and the determination is hereby incorporated by reference for this application.
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| BSEP 15-0101 Enclosure 2 Documentation of Probabilistic Risk Assessment (PRA)
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| Technical Adequacy
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| BSEP 15-0101 Enclosure 2 Page 1 of 110 Documentation of Probabilistic Risk Assessment (PRA) Technical Adequacy Table of Contents 1.0 Overview ........................................................................................... 2 2.0 Basis to Conclude that the PRA Model Represents the As-Built, As-Operated Plant.............................. .................................................................. 2 2.1 BSEP PRA Model History...................................................................... 3 3.0 Identification of Permanent Plant Changes Not Incorporated in the PRA Model ............. 4 4.0 Conformance With ASME/ANS PRA Standard.................................................. 4 4.1 Internal Events and Internal Flooding PRA ................................................... 4 4.2 Fire PRA......................................................................................... 5 4.3 External Events and Shutdown Risk .......................................................... 6 4.3.1 High Winds and External Flooding ..................................................... 6 4.3.2 Seismic ................................................................................... 6 4.3.3 Transportation and Nearby Facility Accidents ......................................... 6 4.3.4 Shutdown Risk........................................................................... 7 4.3.5 Conclusions on External Events and Shutdown Risk ................................. 7 5.0 Methodology to be Used to Assess Surveillance Frequency Changes........................ 7 6.0 Key Assumptions and Approximations................... ......................................... 8 6.1 DC Power Availability and Battery Life........................................................ 8 6.2 Loss of Off-Site Power (LOOP) Frequencies ................................................. 8 6.3 Fire Modeling ................................................................................... 9 7.0 Conclusions on PRA Technical Adequacy....................................................... 9 8.0 References ................. ....................................................................... 9 List of Tables Table 1 - BSEP Peer Reviews - All Open Findings & Observations ............................ 11 Table 2 - BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions....................................................................................14 Table 3 - BSEP Fire Peer Review Findings & Observations Resolutions ...................... 44 Table 4 - BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions .................................................................................. 105
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| BSEP 15-0101 Enclosure 2 Page 2 of 110 1.0 Overview Brunswick Steam Electric Plant, Unit No 1. and No. 2. (BSEP) will follow the methodology provided in NEI 04-10, Revision 1 (i.e., Reference 1), to develop a risk informed Surveillance Frequency Control Program (SECP) for control of Technical Specification surveillance frequencies. NEI 04-10 provides guidance for implementation of a generic Technical Specifications improvement that establishes licensee control of surveillance test frequencies for the majority of Technical Specifications surveillances. Existing specific surveillance frequencies will be removed from Technical Specifications for the affected specifications and placed under licensee control pursuant to this methodology.
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| The NEI 04-10 methodology uses a risk-informed, performance-based approach for establishment of surveillance frequencies and is consistent with the philosophy of NRC Regulatory Guide 1.174 (i.e., Reference 2). Probabilistic Risk Assessment (PRA) methods will be used to determine the risk impact of the revised intervals. PRA technical adequacy has been addressed through NRC Regulatory Guide 1.200, Revision 2 (i.e., Reference 3), which endorses the ASME/ANS PRA, RA-Sa-2009 (i.e., Reference 4). External events and shutdown risk impact will be considered quantitatively or qualitatively as described herein.
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| This enclosure demonstrates the technical adequacy of the BSEP PRA model to be used as the basis for the BSEP SFCP, consistent with the requirements of Section 3.3 and Section 4.2 of Regulatory Guide 1.200, Revision 2 (i.e., Reference 3):
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| * Section 2.0 addresses the need for the PRA model to represent the as-built, as-operated plant,
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| * Section 3.0 discusses permanent plant changes that have an impact on those things modeled in the PRA but have not been incorporated in the baseline PRA model.
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| *Section 4.0 demonstrates that the various technical elements of the BSEP PRA have been performed consistently with the ASME/ANS PRA Standard as endorsed in the appendices of RG 1.200. The peer reviews that have been conducted and the resolution of findings from those reviews are included in Tables 1 - 4. These demonstrate that the pieces of the PRA have been performed in a technically correct manner.
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| * Section 5.0 includes a summary of the methodology that will be used to assess the risk under the SFCP.
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| *Section 6.0 identifies the key assumptions and approximations relevant to the results used in the decision-making process. This section provides assurance that the assumptions and approximations used in development of the PRA are appropriate.
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| 2.0 Basis to Conclude that the PRA Model Represents the As-Built, As-Operated Plant The BSEP PRA Model of Record (MOR) is maintained as a controlled document and is updated on a periodic basis to represent the as-built, as-operated plant. Duke Energy procedures provide the guidance, requirements, and processes for the maintenance, update, and upgrade of the PRA:
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| BSEP 15-0101 Enclosure 2 Page 3 of 110
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| : a. The process includes a review of plant changes, relevant plant procedures, and plant operating data as required, through a chosen freeze date to assess the effect on the PRA model.
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| bThe PRA model and controlling documents are revised as necessary to incorporate those changes determined to impact the model.
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| cThe determination of the extent of model changes includes the following:
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| * Accepted industry PRA practices, ground rules, and assumptions consistent with those employed in the ASME/ANS PRA Standard (i.e., Reference 4),
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| * Current industry practices,
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| * NRC guidance (i.e., References 2-3),
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| * Advances in PRA technology and methodology, and
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| * Changes in external hazard conditions.
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| For plant changes of small or negligible impact, the model changes can be accumulated and a single revision is performed at an interval consistent with major PRA revisions. The results of each evaluation determine the necessity and timing of incorporation of a particular change into the PRA model. An electronic tracking database is utilized to document pending model changes and updates. Previous major PRA model revisions are summarized in Section 2.1.
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| 2.1 BSEP PRA Model History In response to the original Generic Letter 88-20, the 1992 Individual Plant Examination (IPE) was developed to address the risk from internal initiating events including internal flooding. In 1995, this was expanded to create the Individual Plant Examination of External Events (IPEEE) which included seismic events, internal fires, high winds and tornados, external floods, and transportation and nearby facility accidents. The BSEP PRA model has undergone numerous updates and reviews since the original development to maintain a representation of the as-built, as-operated plant in response to improvements in PRA technology and state-of-the-art methodologies. This section presents summaries of the BSEP PRA MOR updates.
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| MOR 2004 addressed Findings and Observations (F&Os) from the 2001 Peer Review along with a failure and unavailability data and success criteria update. MOR 2005 updated the Human Reliability Analysis (HRAs) and answered F&Os from the 2001 Peer Review to support Mitigating System Performance Index (MSPI). MOR 2006 incorporated the ability to cross-tie service air between units. The MOR 2004, 2005, and 2006 revisions did not change methodologies and did not meet the criteria of an upgrade. An independent peer review was performed in 2007 to assess the status of the PRA internal events model compared to the latest PRA Standard and Regulatory Guide 1.200 requirements.
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| MOR 2007 included the implementation of a new diesel room heat-up analysis, the addition of two independent generators to support DC power, and other PRA model improvements consistent with plant design and operation. In 2008, additional minor changes were made and the PRA was re-issued as MOR 2008. MOR 2010 incorporated minor changes to support Regulatory Guide 1.200 requirements and the internal flooding model was upgraded.
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| This PRA model was then subject to a full scope internal events Peer Review conducted by the BWR Owners Group in 2010.
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| MOR 2011 addressed a majority of F&Os from the 2010 Peer Review. Many of the F&Os
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| BSEP 15-01 01 Enciosure 2 Page 4 of 110 were on the recently upgraded Internal Flooding model, for which this effort resolved technical issues and improved the documentation. MOR 2013 incorporated updates to internal events data such as reliability, unavailability, common cause, initiating events, and HRA data. The MOR 2013 scope was limited and as such, it was not considered a PRA upgrade.
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| 3.0 Identification of Permanent Plant Changes Not Incorporated in the PRA Model The current BSEP Model of Record (MOR 2013) is being used for current applications. All permanent plant modifications and ECs that have been implemented since MOR 2013 have been reviewed as part of Duke Energy PRA model maintenance procedures. There are currently no identified permanent plant modification that have a significant impact on the PRA that have not been incorporated into the MOR.
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| 4.0 Conformance With ASME/ANS PRA Standard The following sections describe the conformance and capability of the BSEP PRA against the ASME/ANS PRA Standard (i.e., Reference 4).
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| 4.1 Internal Events and Internal Flooding PRA The following peer reviews have been conducted to ensure the internal events and internal flooding PRA meets the requirements of ASME/ANS PRA Standard:
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| * In 2001, an industry peer review was performed in accordance with Revision A-3 NEI draft "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance" NEI 00-02 dated June 2, 2000.
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| * In 2007, an industry peer review was performed in accordance with Revision A-3 NEI draft "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance", NEI 00-02, dated June 2, 2000. The model was evaluated against ASME-RA-Sb-2005, Addendum B and Regulatory Guide 1.200.
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| * In 2010, an industry peer review was performed in accordance with NEI 05-04 process. The model was evaluated against ASME PRA Standard ASME/ANS RA-Sa-2009 and Regulatory Guide 1.200, Rev 2.
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| Each of the peer reviews evaluated the model for the supporting requirements for internal events, internal flooding, and containment performance (i.e., Large Early Release Frequency).
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| A total of five findings from the latest 2010 peer review that have not yet been resolved are listed in Table 1. All other findings have been reviewed and resolved and can be found in Table 2. These tables provide a complete list of all the findings, resolutions, and dispositions to determine whether the finding has any significant impact on the 5b application.
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| In reviewing the BSEP risk informed License Amendment Request (LAR) for implementation of NFPA 805, the NRC staff evaluated the quality of the internal events PRA model used to support development of the Fire PRA. The objective of the quality review was, "to determine whether the plant-specific PRA used in evaluating the proposed LAR is of sufficient scope, level of detail, and technical adequacy for the application." The results of the NRC staff quality review are documented in the BSEP NFPA 805 Safety Evaluation for transition to a risk-informed, performance-based fire protection program, ADAMS Accession Numbers ML14310A808 (i.e., Reference 5). The staff concluded that the licensee has demonstrated
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| BSEP 15-0101 Enclosure 2 Page 5 of 110 that the internal events PRA model is technically adequate to support the NEPA 805 risk calculation necessary for the license amendment. While this evaluation was not specific to the 5b application, it further demonstrates the technical adequacy of the internal events model.
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| Based on results of the peer reviews and resolutions, the BSEP internal events and internal flooding PRA meets the requirements of the ASME/ANS PRA Standard as clarified by Regulatory Guide 1.200, Revision 2, at an appropriate capability category to support the BSEP Surveillance Frequency Control Program (SECP). The internal events and internal flooding PRA will be used in accordance with NEI 04-10 to assess proposed surveillance frequency changes under the SFCP.
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| 4.2 Fire PRA The BSEP fire PRA was developed using the guidance provided by NUREG/CR-6850 (i.e.,
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| Reference 7) in support of NFPA 805 fire protection program. The fire PRA is built upon the internal events PRA which was modified to capture the effects of fire. In 2012, an industry peer review was performed on the FPRA using NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Guidelines (i.e., Reference 8). The model was evaluated against the ASME PRA Standard ASME/ANS RA-Sa-2009 (i.e., Reference 4), as clarified by Regulatory Guide 1.200 Rev. 2 (i.e., Reference 3). Of the 263 Supporting Requirements that were determined to be applicable, 208 were assessed as meeting Capability Category II or better, 18 were assessed as meeting Capability Category I, and 36 were assessed as Not Met. The peer team also identified one Unreviewed Analysis Method (UAM). Findings were issued for the UAM and for any Supporting Requirement assessed as not meeting at least Capability Category I1. All findings have been addressed either through model changes, additional documentation, or better justification of the basis. Table 3 provides a complete list of all the findings, resolutions, and dispositions to determine whether the finding has any significant impact on the 5b application.
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| The findings and associated resolutions were submitted as part of the NFPA 805 LAR. The results of the NRC staff quality review of the Fire PRA are documented in the BSEP NFPA 805 Safety Evaluation for transition to a risk-informed, performance-based fire protection program, ADAMS Accession Number ML14310A808 (i.e., Reference 5). The quality review concluded that the technical adequacy and quality of the BSEP PRA is sufficient, with the implementation of certain changes (i.e., Implementation Item #13 in Table S-2 of the NFPA 805 LAR), to support risk-informed changes to the NFPA 805 fire protection program. Those implementation items have been incorporated into the Fire PRA, and a focused scope peer review has been conducted, where required.
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| Between September 2014 and May 2015, a focused scope peer review was performed for the incorporation of sensitive electronics in the FPRA and for the Appendix L sensitivity analysis, pursuant to Implementation Items #13.2 and 15, respectively. The peer review was conducted using the general process defined in NEI 07-12, Revision 1, and focused on FPRA technical elements FSS, IGN, FQ, and UNC in ASME/ANS RA-SA-2009, as endorsed by RG 1.200, Revision 2. Within these constraints, the following Supporting Requirements were assessed as not applicable: FSS-A3, FSS-B1, FSS-B2, FSS-C7, FSS-C8, FSS-D9, FSS-E1, FSS-E2, FSS-F1, FSS-F2, FSS-F3, FSS-G1, FSS-G2, FSS-G3, FSS-G4, FSS-G5, FSS-G6, IGN-A2, IGN-A3, IGN-A4, IGN-A6, and FQ-F2. As documented in the peer review report (i.e.,
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| Reference 12), the remaining Supporting Requirements were assessed as meeting Capability Category II or better, with no Findings.
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| BSEP 15-01 01 Enclosure 2 Page 6 of 110 Based on the results of these peer reviews and resolution, the BSEP fire PRA meets the requirements of the ASME/ANS PRA Standard as clarified by Regulatory Guide 1.200, Revision 2, at an appropriate capability category to support the BSEP SFCP. The fire PRA will be used in accordance with NEI 04-10 to assess proposed surveillance frequency changes under the S FOP.
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| 4.3 External Events and Shutdown Risk The following sections describe how external events and shutdown risk are evaluated.
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| 4.3.1 High Winds and External Flooding BSEP developed both the high winds and external flooding PRA models in 2011. In 2012, an industry peer review was performed which evaluated the models against the ASME PRA Standard ASME/ANS RA-Sa-2009 (Reference 4), as clarified by Regulatory Guide 1.200 Revision 2 (i.e., Reference 3). This was a full-scope review of all the Technical Elements of the high wind and external flood events, at-power PRA. All of the external flooding findings from the 2012 Peer Review have been reviewed and resolved and can be found in Table 4.
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| The high winds have been addressed and will be incorporated for use in this application. The resolutions can also be found in Table 4.
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| Based on results of the peer review and resolutions, the BSEP high winds and external flooding PRA meets the requirements of the ASME/ANS PRA Standard as clarified by Regulatory Guide 1.200, Revision 2, at an appropriate capability category to support the BSEP SFCP. The high winds and external flooding PRA will be used in accordance with NEI 04-10 to assess proposed surveillance frequency changes under the SFCP.
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| 4.3.2 Seismic For the IPEEE submitted in 1995, BSEP employed EPRI's Seismic Margins Analysis (0.3g Review Level Earthquake) to identify vulnerabilities to seismic events. In 2014, BSEP completed a Seismic Hazard Evaluation and Screening Report in response to NRC recommendations of the Near-Term Task Force (NTTF) review of insights from the Fukushima Dai-ichi accident (i.e., Reference 11). The results of the review have shown that the IPEEE is adequate to support screening of the updated seismic hazard for BSEP and that the risk insights obtained from the IPEEE are still valid under the current plant configuration. SSCs impacted by frequency changes under the SFCP, therefore, will be assessed against the seismic margins analysis and evaluated in accordance with NEI 04-10.
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| 4.3.3 Transportation and Nearby Facility Accidents For the IPEEE submitted in 1995, BSEP concluded that potential accidents associated with nearby air traffic, runways, roads, railways, waterways, pipelines, and fixed military and industrial facilities are not considered a significant hazard. Structures, systems, and components (SSCs) impacted by frequency changes under the SFCP, therefore, will be assessed against the IPEEE analysis and evaluated in accordance with NEI 04-10.
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| | |
| BSEP 15-0101 Enclosure 2 Page 7 of 110 4.3.4 Shutdown Risk BSEP operates under a shutdown risk management program to support implementation of NUMARC 91-06 (i.e., Reference 10). The shutdown risk management implementing procedure provide guidelines for outage risk management which focuses on proper planning, conservative decision making, maintaining defense in depth, and controlling key safety functions. BSEP will use the shutdown risk management program procedures to assess shutdown risk for proposed surveillance frequency changes.
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| 4.3.5 Conclusions on External Events and Shutdown Risk External hazards screenings have been performed for BSEP to support requirements of the IPEEE and in review of insights from the Fukushima Dai-ichi accident. NEI 04-10 allows for proposed surveillance frequency change evaluations to use hazard screening in the absence of external hazards PRA models. In cases where these methodologies are not appropriate for a surveillance frequency change evaluation, other qualitative or bounding analysis will be utilized to provide justification for the acceptability of the proposed surveillance frequency change. BSEP will follow the NEI 04-10 guidance to assess external event and shutdown risk associated with potential surveillance frequency changes.
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| 5.0 Methodology to be Used to Assess Surveillance Frequency Changes Existing Duke Energy procedures derived from the NEI 04-10 guidance will be used to govern the SFCP and the surveillance test interval (STI) evaluation process. The following steps will be used to assess proposed changes within the BSEP program.
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| *Each STI revision will be reviewed to determine whether there are any commitments made to the NRC that may prohibit changing the interval. If there are no related commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision will proceed. If a commitment exists and the commitment change process does not permit the change, then the STI revision will not be implemented. Only after receiving formal NRC approval to change the commitment will an STI revision proceed.
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| *Systems engineering evaluations and quantitative assessments from available PRA models will be developed for each proposed change. The BSEP internal events, internal flooding, high winds, external flooding, and fire PRAs all meet the requirements for Regulatory Guide 1.200, Revision 2 at appropriate capability categories, and will be used to assess whether an SSC is affected by the proposed STI change. In calculating SSC failure rates, if the breakdown between the standby time-dependent failure rate and the demand-related failure rate probability for affected SSCs is unknown, then the total failure probability will be assumed to be time-related to obtain the maximum test-limited risk condition. The total and cumulative effects on Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) will be assessed, and cumulative risk will be tracked.
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| *If an SSC being assessed is not modeled in the PRA, then appropriate qualitative or bounding risk analysis will be performed for that SSC. Duke Energy procedures derived from NEI guidelines will be used to determine if the qualitative analysis is sufficient for consideration.
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| | |
| BSEP 15-0101 Enclosure 2 Page B of 110
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| *Hazard screening performed for the IPEEE, and review of programmatic assessments performed in response to the Fukushima Dai-ichi accident will be used to assess seismic and transportation and nearby facilities accidents for potential changes in STI. The BSEP shutdown risk management program for implementation of NUMARC 91-06 will be used to assess the shutdown risk.
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| *The results of each STI assessment will be documented and presented to an Expert Panel, referred to as the Integrated Decision-making Panel (lDP). The lOP will normally be the same panel used for Maintenance Rule implementation but with the addition of specialists with experience in surveillance testing and system or component reliability. If the IDP approves the STI revision, the change will be documented and implemented, and will be available for audit by the NRC. If the IDP does not approve the STI revision, the surveillance frequency is left unchanged.
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| *Performance monitoring will be conducted as recommended by the IDP. In some cases, no additional monitoring may be necessary beyond that already conducted under the Maintenance Rule. Performance monitoring helps to confirm that no failure mechanisms related to the revised STI become important enough to alter the information provided for the justification of the interval changes.
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| *The lDP is responsible for periodic review of performance monitoring results. If it is determined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the IDP will reset the STI to the previously acceptable test interval.
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| 6.0 Key Assumptions and Approximations A list of potential contributors to the uncertainty in the PRA was compiled. The list below represents the modeling assumptions and uncertainty that are considered to have the greatest impact on the BSEP PRA results if different reasonable alternative assumptions were utilized. The approaches taken for the assumptions below represent industry best practices and therefore the need for sensitivity analyses will be determined separately for each of the individual surveillance frequency changes evaluated.
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| 6.1 DC Power Availability and Battery Life The DC power system at BSEP is one of the largest contributors to plant risk. Determination of battery depletion times and associated accident sequence timing and related success criteria can potentially have an impact on results.
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| 6.2 Loss of Off-Site Power (LOOP) Frequencies Loss of off-site power initiating events have been shown to be important contributors to plant core damage due to the potential for station blackout and the reliance of many frontline systems on AC power. The LOOP initiator was separated into plant, grid, switchyard and weather induced LOOPs, which allowed the model to apply recovery actions to the higher frequency events (i.e., plant and switchyard). BSEP used generic industry data to calculate
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| | |
| BSEP 15-0101 Enclosure 2 Page 9 of 110 LOOP frequencies. The LOOP frequency has an impact on ODE and Emergency Diesel Generator (EDG) importance.
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| 6.3 Fire Modeling Fire modeling, although following the technical guidance of NUREG/CR-6850, contains several risk important elements that are judged to contain uncertainties for their respective elements of fire risk methodology. These elements include the fire ignition frequency, heat release rates, fire growth curves, fire suppression failure probabilities, severity factors, and Post-initiator human failure event probabilities. While the approaches taken in the BSEP Fire PRA represent the "state of the art" methodology, they are still constrained by the relatively limited data on fire events at Nuclear Power Plants.
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| 7.0 Conclusions on PRA Technical Adequacy The BSEP PRA model is sufficiently robust and suitable for use in risk informed processes such as the Surveillance Frequency Control Program. The peer reviews that have been conducted and the resolution of findings from those reviews demonstrate that the pieces of the PRA have been performed in a technically correct manner. The assumptions and approximations used in development of the PRA have also been reviewed and are appropriate for their application. Duke Energy procedures are in place for controlling and updating the models, when appropriate, and for assuring that the model represents the as-built, as-operated plant. The conclusion, therefore, is that the BSEP PRA model is acceptable to be used as the basis for risk-informed applications including Risk-Informed Technical Specifications (RITS) Initiative 5b.
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| 8.0 References
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| : 1. NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1, April 2007.
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| : 2. Regulatory Guide 1.174, "An Approach for Using ProbabilisticRisk Assessment in Risk-In formed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, U.S.
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| Nuclear Regulatory Commission, March 2011.
| |
| : 3. Regulatory Guide 1.200, "An Approach for Determiningthe Technical Adequacy of ProbabilisticRisk Assessment Results for Risk- Informed Activities," Revision 2, U.S.
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| Nuclear Regulatory Commission, March 2009.
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| : 4. ASME/ANS RA-Sa-2009, "Standardfor Level 1/Large Early Release Frequency ProbabilisticRisk Assessment for Nuclear Power PlantApplications," Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
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| : 5. Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment regarding Transition to a Risk Informed, Performance-Based Fire Protection Program in Accordance with 10CFR 50.48(0) (TAO NOS. ME9623 and ME9624), January, 28, 2015. (ADAMS Accession No. ML14310A808)
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| : 6. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance,"~
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| Revision A-3 draft, June 2, 2000.
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| : 7. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"
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| Volume 2, September, 2005.
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| | |
| BSEP 15-01 01 Enclosure 2 Page 10 of 110
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| : 8. NEI 07-12, Draft Version E, "Fire ProbabilisticRisk Assessment (FPRA) Peer Review Process Guidelines," Nuclear Energy Institute, May 2007.
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| : 9. NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard(Internal Events)," Revision 1 (Draft), Nuclear Energy Institute, November 2007.
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| : 10. NUMARC 91-06, "Guidelines for Industry Actions to Address Shutdown Management,"
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| December 1991.
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| : 11. BSEP 14-0028, "Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident." March 31, 2014. ADAMS Accession No. ML14106A461
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| : 12. EPM Report R2427-0001-01-00, "Focused Scope Peer Review for the Brunswick Fire PRA Against the ASME PRA Standard Requirements," September 2014-May 2015.
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| : 1. BSEP Peer Reviews All Open Findings & Observations________
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| SRIF&O # JFinding S~~~~~Table
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| ]Resolution -
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| 5b Impact In BNP-PSA-034, Attachment 3 Instead of 0.008 for single train, a higher value of 0.01 was There is no describes the Type A human used. For the common cause value of 8E-4, a screening impact to the 5b error screening methodology value of 5E-03 was used. The impact of the higher HEPs is application.
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| employed for the Brunswick negligible with a FV value of 5E-03. The ambiguity in the PRA model. The methodology documentation will be addressed at a later point.
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| follows the screening methodology suggested in NUREG 1792. The screening values used in Tables 1 and 2 contradict the values said to be employed in Section E.3.1 (i.e.
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| 0.008 applied to Type A human errors affecting a single train HR-12 and 0.0008 for common cause (CAT I/Il/Ill) human failure events).
| |
| 2-3 Nonetheless, Tables 1 & 2 in Attachment 3 utilize screening values of 0.01 for single train Type A human errors and 5E-03 for common cause errors. When these screening values were used for the HEPs and the model subsequently quantified, all cross-train Type A HEPs using screening values of 5E-03 having a Fussel-Vesely Importance < 5E-03 are stated to be screened.
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| Table 1. BSEP Peer Reviews - All Open Findings & Observations SR/F&O # Finding Resolution 5b Impact As discussed in BNP-PSA-049, No changes made to the model. Any credit of repair would Meets CAT I, Appendix D, Section D.1, the reduce LERF. Meets CAT I, which provides conservative which provides CET structure allows for the results for LERF. conservative identification of recovery and results for LERF.
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| repair actions that can terminate There is no or mitigate the progression of a impact to the 5b severe accident. This process application.
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| was incorporated into the LEC3TI original analysis, rather than (CAT I) performing a review of 3-12 significant accident progression sequences and then incorporating repair, as would be inferred from the standard.
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| However, it does not appear that significant accident progression sequences were reviewed.
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| There is no evidence that No changes made to the model. Any credit for equipment Meets CAT I, significant accident sequences survivability would reduce LERF. Meets Cat I, which which provides were reviewed to determine if provides conservative results for LERF. conservative LE-C10 engineering analyses could results for LERF.
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| (CAT I) support continued equipment There is no LE-C12 operation or operator actions to impact to the 5b reduce LERF. It was noted that application.
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| (CAT I) this conservative approach with 3-12 respect to equipment survivability was documented in the uncertainty analysis (BNP-PSA-075, Table 1, Item 236).
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| Table 1. BSEP Peer Reviews - All Open Findings & Observations SR/F&O # Finding Resolution 5b Impact BNP-PSA-049, Section 3.1.2 No changes made to the model. Any additional treatment of Meets CAT I, notes that the treatment of scrubbing would reduce LERF. Meets CAT I, which which provides scrubbing by the reactor provides conservative results for LERF. conservative LE-C13 building is treated in a results for LERF.
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| (CAT I) conservative method. This There is no conservative approach was impact to the 5b 3-13 identified in the uncertainty application.
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| analysis (BNP-PSA-075, Table 1, Item 217).
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| There does not appear to be a Standard EPRI PRA codes were used in the analysis. Documenting centralized discussion of Documentation of computer code limitations will be computer code computer code limitations, addressed at a later point, limitations will not Codes used in generic change the references and SAR (and quantified risk SC-C2 associated limitations) are not metrics. There is (CAT I/ll/Ill) discussed. System level no impact to the success criteria for meeting Top 5b application.
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| 6-8 Event gate success criteria are not as well documented in BNP-PSA-033 but appear to be documented on a distributed basis in Heatup Calculations
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| ___________and in System Notebooks. __________________________ ________
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact Unavailability data was taken EDG unavailability was corrected to include all unavailability Resolution of this from MR databases, and outage hours accrued regardless of whether the plant was in finding corrected periods were excluded from the operation or outage condition. In the unavailability value the EDG data to UA calculations (as documented calculation instead of using the unavailability hours while at include all in the BNPUnavail spreadsheet power and dividing them by the total at power hours the accrued that is part of the BNP-PSA-004 value was determined by dividing all unavailability hours unavailability.
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| notebook). In the case of the accrued over the entire window of data collection. The There is no shared EDGs, the BNPUnavail correction resulted in an as expected increase in EDG impact to the 5b spreadsheet is incorrectly unavailability values. application.
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| deleting all EDG QOSs that DA-C1 3 occur when either unit is in an (CAT i/Ill1) outage. Based on the modeling of the shared EDGs, the correct 1-2 treatment would be to NOT exclude any EDG outage (regardless of unit outage condition). This error results in an EDG unavailability that is too low. It should be noted, that based on current data, the impact of this error appears to not be too significant on the results.
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| For SBO, the event tree model An SBO in the PRA model assumes a loss of suppression Resolution of this does not guarantee that a safe pool cooling due to the total loss of AC power on the finding expanded stable state has been achieved, affected unit. There are LOSP cut sets in the PRA results the event trees to AS-A2 The LOSP convolution analysis where power is available and that suppression pool cooling address the (CAT I/Il/Ill) (BNP-PSA-036) includes failed. The recovery of AC power is a subset of those required recovery AC power implicitly. cutsets so that suppression pool cooling is not required to questions for a 1-15 However, the accident be modeled. safe stable state.
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| sequence analysis does not There is no consider the possibility of failure However, for the SBO event tree, there were some non- impact to the 5b of suppression pool cooling conservatism noted, including the no questions involving application.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact following AC power recovery. suppression pool venting or addition actions the operator must perform to maintain long term RCIC operation. The event tree was expanded to address required questions for a safe stable state. The accident sequence notebook was updated (Mar. 2011) to address the modified event tree.
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| Based on the information OSI P1 data was used to make estimates of the component Resolution of this presented in the system standby time for systems with redundant trains. A review of finding collected notebooks (BNP-PSA-062), the the plant specific operating data over a period from 2003 to plant data to PRA estimates the standby time 2008 generally shows that the system operating load is estimate the for components based on the balanced equally across the available trains. In instances component number of trains available vs. where there are two available trains and only one is required standby time for the number of running trains to operate there is a near 50/50 split balance between the redundant trains.
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| DA-C8 (e.g., 50% for a 1 of 2 system). pair. This is similarly seen in the systems with 3 trains with The plant data (CAT I) Actual plant specific data each train carrying approximately 33% of the operating load, supports the concerning standby time is not Based on the results of the data review, in cases where current values collected and evaluated in the there are standby components they are given a standard used. This finding 22 BSEP PRA. However, may be 50% or 33% running time per component as a data is sufficiently available from 0SI Pl to support supported approximation. resolved for the this. SR to be assessed as meeting CAT IlI/ll.
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| There is no impact to the 5b application.
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| For small break LOCA, the high A specific MAAP analysis was performed which verified Resolution of this SB3 end of water break is RCIC as a success path for a 1-inch diameter break. finding performed (CTIl/l) approximately 1" dia., RCIC is a MAAP analysis.
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| (CTII/l) credited for HPI for success, butThrisn no MAAP run was performed to impact to the 5b 1-1 demonstrate the success. application.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O #__ Finding fResolution [ 5b Impact There is no evidence of Resolved in the BSEP Model; the flooding findings have Resolution of this documentation of review or been resolved by improved documentation and PRA model finding improved modification of accident analysis. documentation sequences. Discussions with and PRA model the PRA Staff indicate that IF analysis. This lEs are evaluated using the finding is Transient ET. Specific flood sufficiently impacted equipment is taken out resolved for the of service in conjunction with the SR to be specific flooding IE. assessed as meeting CAT All sequences were added to I/Il/Ill. There is no the general transient event impact to the 5b sequence. No new event trees application.
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| were generated. The IFQU-A1 quantification section of the (NOT MET) flooding report does not discuss how it was determined that 2-9 there were no special flooding sequences that warranted special handling.
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| The Flooding Analysis document RSC 10-05 does not provide information concerning modeling of system/component failures due to pipe failures in that system, as opposed to failure due to flooding and flood propagation. Detailed discussions were necessary to establish the logic by which, for example, Fire Protection Water and Service Water systems can
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact still be credited for injection to the vessel after piping failures in the respective systems.
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| Sequence modifications are not documented. Some fire protection and service water pipe breaks will fail the system as a LPI source by failing the injection path, including consideration of flow diversion effects; this may not be fully accounted for.
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| Operator interview insights are Detailed operator interviews were conducted for the Resolution of this documented in the HRA purpose of confirming procedure interpretations. PRA finding updated Calculator. The information documents have been updated to improve their clarity in PRA documents contained in the HRA Calculator this area. The HFEs mentioned are no longer used in the to improve clarity.
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| was sufficient to demonstrate PRA. This finding is the Capability Category I was sufficiently met. However, the information in The results of operator interviews are put into the HRA resolved for the the HRA did not demonstrate calculator. If there were any special comments from the SR to be that detailed talk throughs with operators, they are included in the operator response tab assessed as HR-E3I Operations and Training for each operator action. meeting CAT IlIlll.
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| (A ) Personnel were conducted for There is no the purpose of confirming impact to the 5b procedure interpretations. For application.
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| example, many of the calculations referred only to an interview conducted with a single operator on 9116-17/2008. A few calculations referred a "talk through" in January 2008, an operator interview on 3/11/2010, or
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution j 5b Impact simulator runs conducted on 1/19/2010. A few calculations (OPER-BLACKSTART, OPER-CNS, OPER-CWSIE) did not
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| - have any input on operator interviews. The purpose and content of these interviews is not evident. Based on the information provided, Capability i/I/ll was not demonstrated __________________________ ________
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| While it was documented that The actions for which there are observed simulator runs are Resolution of this simulator observations and talk- in the Annex section with applicable documentation in the finding added throughs were performed in main body of the calculation to ensure that it can be traced. documentation of most HRA calculations, there is A generic operator discussion sheet was added to the both simulator no evidence that these calculation (BNP-PSA-034). interviews and observations or talk-throughs checklists. This HE4 were used to confirm the The results of operator interviews are put into the HRA finding is (CAT I) response models for the calculator. If there were any special comments from the sufficiently scenarios modeled in the PRA. operators, they are included in the operator response tab for resolved for the For example, there was no each operator action. SR to be interview checklist, assessed as simulator/scenario checklist, or meeting CAT IlI/ll.
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| other documentation to There is no demonstrate that the HRA impact to the 5b analyst confirmed the response application.
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| models.
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| Problems were noted with the OPER-DCDG has been cleaned up and execution errors Resolution of this (CRT-II HRA calculation for OPER- have been added. Additionally the DCDG modification has finding revised the (CAT-G3 DCDG. Specifically, no been revised to eliminate the need to make the connections execution R-3 execution failure probabilities discussed. The HRA has been revised to reflect the change evaluation. There (CTI/l) were assigned to the tasks of in DCDG operation. is no impact to the 3-8 starting and connecting the DG. 5b application.
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| ____________Additionally, the calculation may
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact not have considered all of the necessary breaker manipulations.
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| In general, the HRA calculator All the HFE's have been gone through and information such Resolution of this file was reviewed and found to as training has been added if applicable. If a HEP has no finding added the provide an assessment of the value, (no cognitive procedure or training is not available) missing performance shaping factors then it is documented in the corresponding tab. information to the listed in the SR for the HEP HRA calculator.
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| calculations. Some detail in the There is no calculations could be enhanced. impact to the 5b For example, the operator application.
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| action OPER-LDSHD calculation does not have the HR-G3 cognitive procedure listed and (CAT IlIll1) does not address the training requirements. Calculations for 3-6 OPERMSIVCBP and OPER-DEPRESS1 state that simulator and classroom training are provided but does not provide a frequency. The calculations for OPER-DCDG and OPER-N2SUPPLY do not address training, the cognitive procedure or the staffing requirements.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The only review for initiating A systematic evaluation was performed and documented in Resolution of this events caused by systems the IE Notebook, Calculation BNP-PSA-032, Rev. 9, Section finding performed failure was for mitigating 2.4. Section 2.4 contains a table in which all of the BSEP a systematic systems already modeled. No plant systems are listed and reviewed for their ability to evaluation of each evidence of reviews of other result in a plant trip/initiating event. system as a IE-A5 systems was found. potential IE (NOT MET) contributor. This finding is 4-2 sufficiently resolved for this SR to be assessed as meeting CAT II.
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| There is no impact to the 5b application.
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| The use of a single pipe section The pipe rupture frequencies for Break Outside Resolution of this as the basis for IE frequency for Containment have been revised to incorporate the latest finding breaks outside of containment Pipe Rupture Frequencies from EPRI's Pipe Rupture incorporated the could be nonconservative by a Frequencies for Internal Flooding Probabilistic Risk updated EPRI factor of 100 to 1000 compared Assessments, Revision 2 Report 1021086. Calculation pipe failure to latest EPRI pipe failure BNP-PSA-032, Revision 9 Section 3.2 contains the analysis frequencies into frequency methodology (see using the current methods of calculating pipe break the internal IE-C1 also IE-C6). frequencies by linear foot of piping. The analysis assesses flooding model.
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| (CAT I/Il/Ill) piping systems that can contribute to high energy line There is no breaks outside containment and includes Main Steam Lines, impact to the 5b 1-3 Feedwater Lines, High Pressure Coolant Injection Lines, application.
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| Reactor Core Isolation Cooling Lines, Reactor Water Clean Up Lines and Scram Discharge Volume.
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| The analysis identified the need to include Main Steam Lines breaks as an initiator and Nuclear Task Management
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| #00458348 was generated to include this initiator in future model revisions. The increase in IE frequency of the main _________
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact steam lines was a result of identifying and including additional piping and a common steam chest.
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| The other large BOC initiators analyzed with the new method resulted in un-isolated pipe break frequencies that were near the values of the old EPRI pipe break analysis values.
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| The PRA uses a 82% plant Revised the plant capacity factor to exclude the period from Resolution of this
| |
| -capacity factor. (Section 4.1.2 of 1989 through 1993. The resulting capacity factor of 92.4% finding updated BNP-PSA-032) This appears is based on a 15 year period from 1994 through 2008 and is the plant capacity nonconservative based on the considered more representative of current and future factor to a more IE-C5 past fifteen years. Capacity operations (Section 4.1.2 of BNP-PSA-032 Rev 9). representative factor appears to be over 90% value. There is (CTII) over past 15 years and that no impact to the
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| -1 would be the expectation for the 5b application.
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| future. Use of a 82% capacity factor results in underestimating IE frequencies on a reactor critical year basis.
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| There are questionable The pipe rupture frequencies for Break Outside Resolution of this assumptions on Breaks Outside Containment have been revised to incorporate the latest finding reanalyzed Containment, particularly the Pipe Rupture Frequencies from EPRI's Pipe Rupture pipe breaks use of a nonconservatively Frequencies for Internal Flooding Probabilistic Risk outside EC6 small pipe break frequency for Assessments, Revision 2 Report 1021086. Calculation containment with (OME) Main Steam and Main BNP-PSA-032, Revision 9 Section 3.2 contains the analysis the updated EPRI (OME) Feedwater piping through use of using the current methods of calculating pipe break methodology.
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| 64 old EPRI pipe section frequencies by linear foot of piping. The analysis assesses The analysis did methodology vice consideration piping systems that can contribute to high energy line not change the
| |
| -of piping length. breaks outside containment and includes Main Steam Lines, prior screening Feedwater Lines, High Pressure Coolant Injection Lines, results.
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| Reactor Core Isolation Cooling Lines, Reactor Water Clean Documentation Up Lines and Scram Discharge Volume. The analysis, was added and
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact using the new method resulted in un-isolated pipe break this finding is frequencies that were near the values of the old EPRI pipe considered break analysis values and thus the high energy Break sufficiently Outside Containment initiators are still screened from the resolved for the model. SR to be assessed as meeting CAT 1/11/111. There is no impact to the 5b application.
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| An older EPRI pipe break The pipe break frequency methodology has been update to Resolution of this method based upon piping the EPRI TR-1013141 method. See Table F.15 (RSC finding updated segments is used. This 05) for break frequency categories and Table F.16 (RSC the pipe break methodology had been the 05) for pipe break frequencies. frequency subject of previous F&O IF-D5a- methodology.
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| : 1. The latest EPRI methodology This finding is is based upon piping length, and sufficiently differentiates between whether resolved for the the Service Water is sea water, SR to be lake water, or river water. The assessed as lFEEV-A5 Brunswick Service Water meeting CAT (NOT MET) System is considered a salt I/Il/Ill. There is no water system and thus would be impact to the 5b 6-15 in the category with the highest application.
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| failure frequencies. Note, using as an example 6' Service Water piping, the segmentbased methodology (Section F.8.1) corresponds to a frequency of 5.8E-06/year, which would correspond to about 2 1/2 feet of piping based on the 'spray' failure frequency (100 gpm or less) of sea water service water
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact pipe per EPRI TR-1 013141.
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| Thus, use of the segment-based approach is questioned since it results in underestimating the pipe failure frequency, particularly for the Service Water system which is an important contributor to Internal Flooding scenarios. It is recommended, as it was by the previous F&O, that the pipe failure frequency methodology be updated.
| |
| RS 10-03, Section F.5.1 and The plant specific nature of the potential plant systems Resolution of this F.5., discuss the use of plant impacts on flooding frequency was evaluated based on the finding evaluated specific piping configuration and type of water (fresh, salt, etc.), length of pipe (plant specific the plant specific walkdowns to determine flood - room specific), valve rupture and maintenance induced nature of potential initiating event frequency. floods was evaluated in Section F.8 of RSC-1 0-05. plant systems Additionally, generic failure impacts on rates were used for pipe break The BSEP historical flood data is included in section F.1 .1 of flooding IFEV-A6 frequency. RSC-10-05. The EPRI pipe failure rate used BSEP data frequency. This (CAT I) where appropriate to determine the pipe failure rates used finding is It did not appear that plant- (EPRI TR-1 013141 - reference 145 in RCS-1 0-05). sufficiently 3-14 specific information was resolved for the gathered with respect to the SR to be flood LIKELIHOOD. Specifically, assessed as there was no operating meeting CAT IlIll1.
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| experience related to water There is no hammer or material condition of impact to the 5b fluid systems at Brunswick. application.
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| | |
| - Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact Documentation of SR IFQU-A2 Table F.B.1 maps flooding initiating events to impacted Resolution of this does not provide a link between rooms. Tables F.5 and F.6 of RSC-10-05 map components finding mapped the flooding IE and the to flood zones (rooms). Table F.14 of RSC-10-05 maps the flood sources, equipment failed for that specific propagation of flood through the flood zones (and rooms). flooding initiating flooding initiator. By inspection events, and flood of the model, equipment and propagation to the IFQU-A2 system fault trees have been appropriate modified to include flood- rooms.
| |
| (NOT MET) induced failures. No finding isThis 1-2 documentation of the events or sufficiently systems impacted is included in resolved for the the flooding analysis SR to be (maintained as a separate assessed as spreadsheet, see IFQU-B1). meeting CAT Thus, it is not possible to fully I/ll/1l1. There is no verify if the modifications were impact to the 5b done correctly, application.
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| The analyses of all HRA are 1) The flooding HRA's now use a CBDTM method, Resolution of this documented in the HRA which is used for other BNP specific HRA's. finding converted Calculator and were not readily 2) The CBDTM method was used, and alarms were the flooding HRAs available for implemented. System alarms were integrated to the CBDTM review. Based on the time sparsely under the consideration that the leak may method, available for HRA Calculator not be severe enough to reach the setpoint. Sump implemented
| |
| ]FQU-A5 review, the following issues and alarms were used in all cases.,a alarms, and (NOT MET) comments are 3) Cues are listed in the HRA calculator, "cues" ta. implemented and provided. This should not be 4) Operator discussion was implemented and documented 2-0 considered an all inclusive list documented per BNP-PSA-034 operator given the time restraints on the Time available is presented in tables F.A. 1, F.A.2, F.A.3, discussions. This review, and F.A.4. For events with spray, the analysis does not finding is credit any HRA's for mitigation or recovery for equipment in sufficiently While the HRA analyses did the path of the water spray, resolved for the include many important factors SR to be used to determine a human assessed as
| |
| _______error meeting CAT
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding [Resolution 5b Impact probabilities (HEPs) associated I/li/Ill. There is no with isolating flood events, the impact to the 5b following issues were judged to application.
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| potentially impact the calculated HEPs:
| |
| a The methodology used for calculating the HEPs for the flooding termination was the annunciator response methodology. There was no discussion in the flooding documentation regarding the use of, or acceptability of the use of, this methodology. Flooding HRA documentation should be enhanced to include a discussion of the methodology, and the justification thereof, for calculating HEPs.
| |
| * The use of the annunciator response methodology did not appear to account for all expected annunciators. For example, for large turbine building floods, it does not appear reasonable that only one annunciator would be received.
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| Under the large turbine building flood conditions, it would be o
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding [Resolution 5b Impact expected that the loss of circulating water pumps and the plant transient would cause multiple alarms.
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| Therefore, this HEP could be significantly under-estimated as the annunciator response methodology modeling only assumed that one alarm would be received.
| |
| * Since the detection and diagnosis of a flooding situation is critical to flood mitigate, a detailed discussion of the detection systems and alarms available for diagnosis is necessary to ensure a realistic HRA. Neither the flooding documentation, nor the HRA calculator files, contains a complete description of the cues available to the operator for flood detection.
| |
| Therefore, the accuracy of the HEP for flood termination is questionable.
| |
| * It does not appear that operator interviews or talk-throughs were used to determine the floodincj
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution ] 5b Impact HEPs. Interviews with operators and their response to various alarms would provide invaluable insights into the flood response and would provide for a more realistic approach.
| |
| *It is not clear that the "time available" value used in the HEP was appropriate for some scenarios. The time available used was the time to which the flood would reach the height of critical equipment.
| |
| However, the time of the flood to reach the flood isolation valves, in some scenarios, may be time limiting. Additionally, the impact on timing if damage is due to spray, vice submergence, must be considered. The analysis should be more specific regarding which valves need to be isolation, their locations, and the ability to isolate before the flood scenarios cause them to be unisolable.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact No consideration is documented BNP-PSA-035 section F.13 discusses the impact of flooding Resolution of this for the impact of flooding on on the internal events HRA. finding operator actions modeled in the documented the HRA which require operator BNP-PSA-034 and BNP-PSA-035 discuss the flooding impact of flooding actions in areas subject to alarms associated with credited operator actions. Timing, on internal flooding. There is no discussion including discovery, was broken down to the limits events, flooding in the documentation of how the presented in BNP-PSA-035 section F.13. alarms associated flood would be discovered and with credited thus no discussion of the time it BNP-PSA-035 section F.9, discusses the effects of the flood operator actions, IFQU-A6 would take plant personnel to on mitigation equipment. and the effects of (NOT MET) start to respond. There is no the flood on documentation or discussion of mitigation 6-13 this in Section F.11. No equipment. This discussion of whether or not the finding is mitigating equipment listed in sufficiently table F.20 would be affected by resolved for the the flood itself. SR to be assessed as meeting CAT I/Il/Ill. There is no impact to the 5b
| |
| ____________application.
| |
| Direct effects were included. Impacts of spray are discussed in section F.10 of RSC Resolution of this The only indirect effect included 05. Equipment considered for spray impact is listed in Table finding discussed was submergence. F.21 of RSC-10-05. Effects of High Energy Line Break the impacts of (pressure, humidity, condensation, temperature and pipe spray events and I FQU-A9 whip) are discussed in section F.1 1 of RSC-10-05. high energy line (NOT MET) breaks. This finding is 1-33 sufficiently resolved for the SR to be assessed as meeting CAT
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact I/Il/Ill. There is no impact to the 5b application.
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| There is no evidence that this The impacts of internal flooding on large early release Resolution of this requirement was addressed. frequency were evaluated during the LERF cutset reviews finding analyzed IFQU-A10 The flooding analysis did not performed as part of Revision 10 of the Quantification the impact of (NOT MET) appear to analyze any impact Calculation (BNP-PSA-030). The review found that internal flooding on the on the LERE model due to the flooding was a contributor to LERF and that the flooding LERF model.
| |
| 6-14 flooding initiators, hazards were being captured appropriately in the model. There is no impact to the 5b application.
| |
| The report RSC 10-05 does not Table F.B.1 maps flooding initiating events and source pipes Resolution of this contain the information that lists to impacted rooms. Tables F.5 and F.6 of RSC-10-05 map finding mapped what equipment is damaged for components to flood zones (rooms). Table F.14 of RSC the flood sources, each pipe failure scenario, thus 05 maps propagation of flood through the flood zones (and flooding initiating there is no documentation rooms). events, and flood available for review for this propagation to the IFQU-B2 aspect of defining the plant appropriate (CTIl/l) equipment subject to damage rooms. This (A IIl) from each flooding initiator. This finding is 6-0 information may be contained in sufficiently the database used in performing resolved for the the flooding analysis, but it SR to be could not be confirmed. assessed as meeting CAT 1/11/Ill. There is no impact to the 5b application.
| |
| IFSN-A2 The flooding analysis (as Flooding analysis in the HRA Calculator implement the Resolution of this (NOT MET) documented in RSC 10-05) alarms used to tell operators that there is a flood occurring finding included discusses sumps and drains, in the system. These alarms are in the control room the use of
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact 1-25 curbs, spray shields, and (described in BNP-PSA-035) and redundant alarms are flooding alarms in watertight doors. No mention is available on the Radwaste panel, there is sufficient alarms the HRA made of flood alarms (or other for the operator. Turbine building floods with operator calculator. This information that would alert actions do not appear in the higher cutsets, and pose finding is operators of the flood), blowout negligible risk. sufficiently panels or HVAC dampers. As resolved for the flood screening is performed SR to be based on assumed operator assessed as intervention, the lack of meeting CAT information concerning alarms I/Il/Ill. There is no impacts the ability to accurately impact to the 5b assess the probability of application.
| |
| successful flood termination RSC 10-05 does not identify any Alarms are documented in the HRA calculator for the floods Resolution of this automatic flood isolation that have operator actions. Isolation features and timing are finding included features, No operator discussed in BNP-PSA-035. the use of indications are discussed. flooding alarms in However, the qualitative the HRA screening of flood areas calculator. This evaluates successful operator finding is IFSN-A3 action to isolate any flood sufficiently (NOT MET) assuming that the flood is resolved for the immediately detected. The flood SR to be 1-20 documentation needs to indicate assessed as how each flood would be meeting CAT detected and the time to identify I/Il/lll. There is no the occurrence of the flood impact to the 5b needs to be factored into the application.
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| HEP calculation. As the current evaluation is non-conservative,
| |
| ____________this SR is assessed as not met.____________________________
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # ] _Finding ______[Resolution 5b Impact _
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| The methodology is considered Impacts of spray are discussed in section F.10 of RSC Resolution of this sufficient and-robust for 05. Equipment considered for spray impact is listed in Table finding assessed identifying flood damage due to F.21 of RSC-10-05. the impact of submergence. EPRI TR- spray events and 101 9194 recommends that at Effects of High Energy Line Break (pressure, humidity, high energy line least a 10 foot radius from a condensation, temperature and pipe whip) are discussed in breaks. This pipe be considered for impact of section F.1 1 of RSC-10-05. finding is spray due to pipe failure. Many sufficiently industry PRAs assume up to 30 resolved for the feet for spray impact. While SR to be spray due to water falling down assessed as propagation pathways has been meeting CAT I/I1.
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| assessed for impact (or lack There is no thereof) on SSCs, there is no impact to the 5b IFSN-A6 evidence that this was done for application.
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| (NOT MET) any equipment. Equipment in flood zones such as upper 6-16 levels of the reactor building would likely be subject to this failure mechanism.
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| The R.G. 1.200 qualification for this SR requires that a qualitative assessment of pipe whip, jet impingement, humidity, condensation and temperature concerns in the flooding analysis. The current flooding documentation does not provide this assessment.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact Section F.1.3 of RSC10-05 Section F.4.8 provides basis for not considering drain paths, Resolution of this states that flows through drains wall penetrations, cable trays, and HVAC ducts as they are finding provided a were considered, but there is no insignificant compared to the flood propagation through basis for not discussion concerning drain door, stairwells, and gratings, considering some paths as a possible propagation potential path between rooms in the propagation Auxiliary Building basement nor pathways. This IFSN-A8 the basis for not considering finding is (CAT I) such paths. Discussion of drain sufficiently paths not included, Propagation resolved for the 1-26 through wall penetrations, cable SR to be trays, and HVAC ducts do not assessed as appear to be considered. As meeting CAT I1.
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| only propagation through doors, There is no stairwells, and gratings are impact to the 5b considered, the requirements for application.
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| Cat II are not met.
| |
| While multi-unit flooding was F.2.1 .2 dismisses the propagation of floods in the turbine Resolution of this considered, all scenarios were building due to "massive" amount of water required to fail finding improved
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| -screened out on the basis of the doors between zones. Section F.1.3 assumes that the flooding physical barriers that are roll up doors in the turbine building are pushed open once documentation to assumed to prevent flood the water has reached 4 inches on the 20ft elevation, address multi-unit ISFN-A1 1 propagation between units. flooding (NOT MET)
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| | |
| ==Reference:==
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| BNP-PSA-035, Rev. 4 scenarios. This IFEV-A4 However, the turbine building finding is (NOT MET) has complete communication sufficiently between unit 1 and unit 2. resolved for the 1-27 Therefore, water can spread SR to be from one unit to other and vice assessed as versa. The flood scenario is meeting CAT described more detail in Section I/lI/Ill. There is no F.2. 1.2 of the notebook. In impact to the 5b addition, turbine building roll-up application.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact doors on elevation 20' do not appear to have been considered
| |
| - as a propagation path between units. So, it appears that the screening of all inter-unit turbine building floods may not be correct.
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| Brunswick Flooding PRA credits Drains are not credited as mitigation paths in RSC-1 0-05. Resolution of this operation of drains for mitigating Drains are mentioned as a possible mitigation path for the finding removed flooding events. Per EPRI TR- EDG rooms, but that is only listed as one of the possible credit for drains 1019194, there is wide paths which included doors and stairwells, as flood variability in modeling on this mitigation paths.
| |
| IFSN-A13 issue, with widespread industry This finding is (NOT MET) practice of not crediting the sufficiently functioning of drains due to the resolved for the 6-18 high probability of sump pump SR to be failures and clogging of drains, assessed as There is no discussion in RSC meeting CAT 10-05 concerning the reliability 1/11/111. There is no of the drains as a mitigating impact to the 5b
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| _______system for flooding, application.
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| Flood sources were screened if Flooding alarm locations are discussed in BNP-PSA-034 Resolution of this more than 8 hours is required to and BNP-PSA-035. finding included reach a one foot depth in an For flooding initiators where operators have time to isolate the use of area. However, there is no the flood, indications are discussed with their respective flooding alarms in IFSN-A14 discussion of whether indication entry in the HRA Calculator. the HRA (NOT MET) is available to identify the flood calculator. This or if isolation can be performed finding is 1-28 given the flooding is occurring, sufficiently resolved for the SR to be assessed as meeting CAT I1.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact There is no impact to the 5b application.
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| Table F.5 of RSC 10-05 lists all Information has been corrected, flooding analysis has been Resolution of this of the potential flooding sources, updated. Fire protection piping described in Table F.3 of finding corrected however one of the major RSC-1 0-05. Pipe break frequencies for Fire Protection Pipe information assumptions is that all fire (by Pipe Identification Number) are included in Table F.16 of concerning the protection sprinkler systems are RSC-1 0-05. fire protection IFSO-A1 dry-type systems.SD-41 system. This (NOT MET) contains a listing of fire finding is protection systems and most sufficiently 1-1 are wet type systems. Since the resolved for the inclusion of fire protection could SR to be significantly alter the screening assessed as and scenario development, this meeting CAT is considered to be not met. 1/11/Il1. There is no impact to the 5b application.
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| Pipe breaks and valve body Gasket and expansion joint caused flooding is now covered Resolution of this failures were the only failure in section F.12 of RSC-10-05. Human induced mechanisms finding added mechanisms identified. Although are covered in section F.8.3 of RSC-1 0-05. gasket and plant experience includes a expansion joint IFS O-A4 gasket failure, failures of caused flooding (N_ E) gaskets, expansion joints, etc. as well as human are not discussed. Therefore,inue SR is not met as all required mechanisms to 1-2 mechanisms are not the model. This considered. finding is sufficiently resolved for the SR to be
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact assessed as meeting CAT 1/11/111. There is no impact to the 5b application.
| |
| The temperatures and Section F.6 of RSC-10-05 describes the characteristics Resolution of this pressures of condensate, (temperature and pressure) of the various flood sources. finding adequately feedwater, and nuclear service Volumes of flood sources provided in Table F.9 of RSC-1 0- described the piping are not considered. The 05. Flood flow rates provided in Table F.12 of RSC-10-05. characteristics of flow rates for circulating water flood sources.
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| IFSO-A5 piping are based on one pump Circulating water piping based on 4 pump flow - see Table This finding is (NOT MET) instead of all four. This SR is F,12 of RSC-10-05, sufficiently judged to be not met. resolved for the 1-23 SR to be assessed as meeting CAT I/1l/Ill. There is no impact to the 5b application.
| |
| Most of the level of detail Table F.12 maps flood sources to room and Table F.B.1 Resolution of this required to relate the flood maps flooding initiating events to impacted rooms. Tables finding mapped sources to actual rooms and F.5 and F.6 of RSC-1 0-05 map components to flood zones the flood sources, piping and components is (rooms). Table F.14 of RSC-10-05 maps propagation of flooding initiating located in a vendor database flood through the flood zones (and rooms). events, and flood (NSOTMET that is not part of the propagation to the (FNOTAET documentation file. Therefore appropriate (NOT MET) this SR is not met. rooms. This finding is 1-24 sufficiently resolved for the SRs to be assessed as meeting CAT
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| _________________________________I/Il/Ill. There is no
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b impact impact to the 5b application.
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| Documentation of the internal IFSN-B1 - Discussion of flooding scenarios enhanced in Resolution of this flooding work was not conducive section F.4 of RSC-1 0-05. finding improved to supporting PRA applications, documentation maintenance and upgrade, or IFSN-B2 (c) - Assumptions for submergence, spray, and the PRA peer review. Much of the temperature and other flood effects given in sections F.1 .3, model analysis.
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| information was contained in a F.9, F.1 0 and F.1 1 of RSC-1 0-05. This finding is IFSN-B1 contractor flood database that sufficiently (NOTI MET) was not captured in the IFSN-B2 (d) - General screening criteria for analysis in resolved for the IFSN-B2 Brunswick document system sections F.1.2 and F.1.3 of RSC-10-05 and specifics are SRs to be (NOT MET) and which was not accessible given in greater detail in sections F.2 and F.3 of RSC-10-05 assessed as IFSO-B1 for review, on an area by area basis. meeting CAT (NOT MET) There is no I/Il/Ill.
| |
| IFEV-B1 As noted in IFSN-B1, the flood IFSN-B2 (e) - Flooding scenarios considered are identified impact to the 5b (NOT MET) scenario development in section F.2 of RSC-1 0-05. Screening of flooding application.
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| IFEV-B2 documentation is not believed to scenarios is performed in section F.3 of RSC-10-05.
| |
| (NOT MET) provide all of the information Retained flooding scenarios are described in section F.4 of IFQU-B1 needed to fully describe the RSC-1 0-05.
| |
| (NOT MET) scenario development process.
| |
| IFQU-B2 In addition to other items noted IFSN-B2 (f) - Pipe break, spray and human induced (NOT MET) in the IFSN-A SRs, items initiators were added to the model, Tables F.38 and F.39 of (c),(d),(e), and (f) of SR IFSN- RSC-10-05 identify components that were modeled with 1-31 B2 need to be included as part flooding failures. The flooding initiators are listed in annex of the documentation. Also, a F.C of RSC-1 0-05.
| |
| listing of the specific components assumed to be Tables F.5 and F.6 of RSC-1 0-05 identify flooding failed in each flood area needs components by flooding zone and Tables F.38 and F.39 of to be provided. Discussion of RSC-1 0-05 identify PSA components that were modeled propagation pathways could with flooding failures. Flooding pathways are described in
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # ]Finding IResolution 5b Impact have been at a greater level of section F.4 of RSC-1 0-05 for important flood areas. Flood detail. propagation on room by room basis is provided in Table F.14 of RSC-10-05. Assumption on behavior of doors is Documentation of risk-important provided in section F.1.3 of RSC-10-05.
| |
| components within each flood area appeared incomplete in the IFQU-A5 - Section F.13 of RSC-10-05 gives screening HRA walkdown notes. Documentation for flooding events to determine importance of flooding of characteristics of flood events. Tables F.26 and F.27 identify operator actions that pathways between zones was would be impacted by flooding events. Additional HRA weak and/or incomplete, often development for flooding scenarios is included in BNP-PSA-indicating only the first pice of 0034.
| |
| equipment to fail as flood waters rise. The level of detail in the IFQU-A1 - System failures due to piping failures given in walkdown notes was considered section F.9 and Tables F.19 and F.20 of RSC-10-05.
| |
| weak. Documentation and discussion of flood pathways other than doors was limited.
| |
| Assumptions on door behavior, such as pressures doors would withstand or flood propagation rates through door gaps, were not documented.
| |
| Operator actions to be modeled in the Human Reliability Analyses for Flooing were not clearly identified and described, including documentation of alarms that would be considered for specific flooding scenarios.
| |
| Review of plant operating experience pertinent to these HRA analyses were not documented. It was not possible
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact to verify proper development of the HRA Calculator Files, as discussed under IFQU-A5. This contributed to being unable to verify that IFQU-A5 requirements were met.
| |
| As discussed in IFQU-A1, the Flooding Analysis does not provide information concerning modeling of system/component failures due to pipe failures in that system, as opposed to failure due to flooding and flood propagation.
| |
| Documentation of the general philosophy and approach used for the flooding analysis can be improved.
| |
| Parameter values were selected 1) LOSP curves were updated to more modern composite Resolution of this with regards to the PRA curves. Appendix C tables 4.7-2, 7.7-5. finding updated Standard's requirements for HR 2) Component failures have been updated in the LOSP curves and and DA. Consideration of documentation. See BNP-PSA-049 Appendix C component severe accident conditions upon failures. There is LE-E1 these parameters is provided in no impact to the (CAT I/Il/Ill) Appendix M, or in some 5b application.
| |
| instances Appendix C, of the 1-19 BNP-PSA-049 notebook.
| |
| Section G of LE notebook captures the human error modeling, and incorporated the general methodology approach
| |
| ___________used in Level 1 HRA.
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact However, the data values documented in BNP-PSA-049 were developed during a previous PRA update. It appears that some values may need to be updated to be consistent with changes in the Level 1 data. For example, aSP recovery values (such as ACP1XHE-MN-OFFE) are not consistent with the current OSP recovery curve (and LOSP is now categorized by type of OSP failure as opposed to a composite value). On the other hand, changes in component failure data appear to have been updated in the Level 2 trees.
| |
| However, the documentation does not indicate that the values shown in BNP-PSA-049 have been superseded.
| |
| There is very limited discussion Uncertainties are discussed in detail in BNP-PSA-075. Resolution of this of the impact of variability / Table 1, items 228 to 271. This outlines both the finding discussed sensitivity in time to core uncertainties associated with the LERF basis and the the uncertainties LE-G5 damage amongst different limitations for use, associated with (NOT MET) methodologies upon potential the LERF basis applications in Appendix C of and the limitation BNP-PSA-049. While limitations for use. This 6-2 of the quantification process are finding is discussed in BNP-PSA-030 sufficiently Sec.3.6, that discussion is not resolved for the
| |
| _______pertinent to SR LE-G-5. It is SR to be
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact concluded there is not sufficient assessed as discussion of the limitations of meeting CAT the LERF analysis that could I/Il/Ill. There is no impact different applications, impact to the 5b thus this documentation application.
| |
| __________requirement is NOT MET.__________________________ ________
| |
| Dependency analysis was Combinations of operator actions among the top 95% of Resolution of this performed on the identified HFE cutsets were analyzed, and the larger bulk of dependencies finding analyzed combinations (see BNP-PSA- addressed to ensure a correct value. See BNP-PSA-034, combinations of 034 and associated section E.4.1 .5 operator actions spreadsheets). The dependency and assessment approach used dependencies to appears to be appropriate. In ensure correct developing recovery rules to be values. There is applied to the cutsets, maximum no impact to the combinations of 3 HFEs were 5b application.
| |
| included. Any cutsets with greater than three HFEs that QU-C2 meet the recovery rule criteria are recovered to a minimum joint HFE of 1 E-6 (and often 3-9 higher). As a result, there are cutsets that contain more than three HFEs that are being recovered to a higher frequency than may be warranted (either because one or more of the additional HFEs may be independent of the others, or because the joint HFE probability is still above the floor value of 1 E-6 (and often higher).
| |
| As a result, there are cutsets that contain more than three
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact HFEs that are being recovered to a higher frequency than may be warranted (either because one or more of the additional HFEs may be independent of the others, or because the joint HFE probability is still above the floor value of 1E-6 and hence could be reduced further). This conservatism appears to increase the calculated CDF/LERE by at least a modest amount.
| |
| It is stated in BNP-PSA-030 that The initial cutset review is documented in Section 3.7 of Resolution of this the top 200 cutsets have been BNP-PSA-030. The cutset reviews are completed by a finding reviewed a reviewed. However, to panel of PRA personnel. The full range of cutsets was sample of cutsets determine if there are logic reviewed strategically to understand the most significant over the full range problems buried deeper down in cutsets, ensure that the results are reasonable and of significance the cutsets, A sample should be complete. The panel performs the review by analyzing the and improved QU-D1 taken of significant cutsets or first 200 cutsets in detail, analyzing approximately 100 documentation.
| |
| (NOT MET) sequences. The correct cutsets from the 201 through non-significant at random This finding is definition of significance is intervals, and analyzing a small number of the non- sufficiently 2-7 stated in the quantification significant cutsets for completeness. Accident Sequence resolved for the notebook. The review should cutsets were also reviewed to ensure cutsets are correct SR to be include a sampling of cutsets or and consistent with system model and success criteria and assessed as sequences over the full range of to ensure significant and non-significant cutsets are showing meeting CAT significance (i.e., top 95% of up appropriately. Section 3.3 and 3.4 provide a detailed I/Il/Ill. There is no cutsets contributing to review and description of significant cutest (i.e. individually impact to the 5b DF/LERF).
| |
| C______ contributing more than 1% to CDFI/LERF). application.
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact Section 3.9 of BNS-PSA-030 Revised the BSEP Quantification calculation BNP-PSA-030 Resolution of this compares BNS PSA results to include an enhanced similar plant review. Section 3.9 finding revised the against other units. However, contains two tables, one that compares very broadly quantification there is no analysis of Brunswick 1 &2 against other BWRs and another that calculation to contributors as required for Cat compares Brunswick 1&2 to Hatch and Browns Ferry in include the I1.In addition, no references are more detail. The broad review looks at analysis of provided for the other PRAs that differences/similarities in plant design, ODE and LERF. The contributors and are compared to (e.g., are these in depth review examines the individual initiators to COF. references for the QU-D4I IPE results, or current results, Most of the Brunswick model initiators are within an order of review of similar (A ) etc.) magnitude and are considered to compare similarly. The plant PRAs. This 4-7 largest outlier in the comparison is the contribution of internal finding is flooding initiators to the Brunswick core damage frequency. sufficiently The large difference is due to a Brunswick specific design resolved for the characteristic were a break in a condensate pipe traveling SR to be through the cable spread rooms causes an internal flood that assessed as fills the room before mitigation activities can be successful. meeting CAT IlIll1.
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| There is no impact to the 5b application.
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| Failures of the F032A and 1-B21-F032A/B are stop check valves that are upstream of Remains F032B outboard feedwater the F010OA/B. The likely hood of two check valves failing in unmodeled due to check valves are not modeled, series is improbable. During HPCIIRCIC operations in the low probability.
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| The system notebooks state plant, these valves are shut by MOV thus sealing them into HPCI (F032A) or that such failure is considered the closed position unlike normal check valves which may RCIC (F032B)
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| SY-A13 "improbable". However, the stick in the open position due to fouling of the hinge system (CAI/l/ll) model does include the failure of mechanism. The PRA models them with a fails to remain unreliability is two inboard check valve F010OA to open BE due to the credit given for feed in HPI.oresf 4-5 reopen. magnitude higher than the check valve failure probability. The unmodeled failure mode is bounded by the uncertainty
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding jResolution 5b Impact of the system unreliability.
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| There is no impact to the 5b application.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # [Finding JResolution f 5b Impact Generic MSO scenario 2e The identified MSO scenarios were re-evaluated using Resolution of this appears to be inadequately deterministic and thermal hydraulic methods for individual finding re-evaluated dispositioned. The scenario MSOs and combinations of MSOs. The results of the re- the MSO scenarios identified in NEI 00-01 is a drain evaluation concluded that the individual MSOs and mentioned. These of the vessel, while the rough combinations of the MSOs did not result in a failure of MSOs remain calculations evaluate this as credited components, addition of new initiating events or screened from essentially (word in Attachment 3 a change in accident sequences. These MSOs remain as inclusion in the EPRA of the component selection screened from inclusion in the FPRA model. The analysis model. There is no report) a depletion of the of these MSOs has been updated in Attachment 3 of the impact to the 5b ES-B2 suppression pool. The loss is component selection calculation, BNP-PSA-085. application.
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| (CAT 1,1II) estimated as a 200 gpm loss, ES-D1 (CAT 1/l11/1ll) which can be an issue either a) long term for inventory, or b) in PRM-B9 combination with other small (CAT I/Il/Ill) losses (See NEI 00-0 1 for PRM-C1 guidance on combining MSOs).
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| (CAT I/WI/ll)
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| Generic Scenario 2d appears to also be a possible long-term 1-2 issue (with multiple seal failures),
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| or an issue in combination with other small losses.
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| Scenario B21-2c (Main steam drain line) includes an evaluation of flow size listed as 0.03 square inches based on a single flow path. However, multiple drain line openinqs are possible.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The MSIVs spurious operation Given the MSIVs are normally open during power Resolution of this appears to be modeled as a operation, MSIV spurious opening or failure to close finding added two new ES-A5 failure of containment isolation cannot be a fire-induced initiating event. MSIV LOCA accident (CAT I, II) under gate 1 Si. This spurious sequences. There is ES-A6 operation does not appear to be However, two new MSIV LOCA accident sequences were no impact to the 5b (CAT I, II) modeled as either an initiating created to model a fire-induced post-trip MSIV spurious application.
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| ES-B2 event or LOCA, or showing to opening or failure to close (MSO-B21-2b). These (CAT I, II) impact RCIC/HPCI operation, sequences do not credit HPCI or RCIC and include the ES-D1 loss of the condenser.
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| (CAT I/Il/lll) Most BWR FPRAs include MSIV failure to close or spurious re- This has been documented in Section 3.3.1.4 and opening as a large or medium Attachment 3 of the component selection calculation 1-6 LOCA, given downstream BNP-PSA-085.
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| opening of TBVs or other large steam line valves.
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| The FPRA modeling does not A detailed review of the fire induced initiating events was Resolution of this
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| -include mapping of multiple point performed, with particular attention to those initiating finding performed a estimate initiating events to events identified by the Peer Review, and was detailed review of fire ESA1 specific equipment. This includes documented in Section 3.3.1.4 and Attachment 8 of the induced imitating E-l the following lEs: Loss of offsite component selection calculation BNP-PSA-085. The events. Logic for fire (NOT MET) power, Inadvertent opening of review found that all initiating events had been adequately induced LOOP was ES-A2 SRV (%1"1-S), Loss of DC Power addressed except for fire induced LOOP. Based on the added to fault tree (CAT I/Il/Ill) (%1TDC1A1, %1T_DC1B2), sensitivity study cited in F&O 1-8, no change was made where appropriate.
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| ES-A3 Loss of Switchboard for the loss of DC power, because DC initiators are This finding is (NOT MET) (%1T_DC1A, %1TDC1 B), and adjacent to response model events in the fault tree. Logic sufficiently resolved FQ-A2 loss of AC Bus (%1TEEl, for fire induced LOOP was added to the fault tree where for both SR ES-Al (CAT I/1l/Ill) %1TE_E2). In essence, these are appropriate. Inadvertent SRV opening was removed from and SR ES-A3 to be treated as a plant transient (in this IAN"*G005 which is present under IAN^G178 as assessed as meeting case, an MSIV closure event) documented in Rev 2 of the change log, Attachment 9 of CAT I/ll/Ill. There is 1-8 followed by a subsequent failure BNP-PSA-085. Attachment 3 of the component selection no impact to the 5b of the equipment. calculation BNP-PSA-085 documents MSOs that were application.
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| evaluated as possible initiators but determined not to be A sensitivity case was requested creditable.
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| for the loss of DC Power Al and
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # {Finding jResolution j 5b Impact ioss of offsite power to determine the possible impact on the COD P.
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| The results show some differences in the cutsets and the CCDP results, mainly due to actuation logic (applied under the IE logic), restart logic, and failure of CRD. Overall, the CCDP following the IEs is slightly higher than assuming the subsequent failure of the equipment.
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| Significant tracing was performed of the logic for each IE. In most cases, the IE logic was ORed with the equipment logic.
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| However, there were exceptions.
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| For example, for the inadvertent SRV opening; gate IAN1 G178 (HEADER A ISOLATED AND NOT RECOVERED) included the IE but not the equipment logic for SRV opening. Another example:
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| Gate #U13 (S2 LOCA OR SORV WITH ONE OR MORE SRVS FAILING TO RECLOSE) includes SRV logic above, but only for 2 or more SRVs. As a result, the single SRV opening for the IE is not included under this logic.
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| Review of LOOP logic indicated several locations where consequential LOOP was not
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact included; although the logic included in most cases other fire logic such as MSIV closure, or other assumed fire lEs.
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| Feedwater and HPCI overfeed is The applicable impact of Feedwater and HPCI overfeed, Resolution of this not included in the FPRA as initiator events, is already appropriately modeled. finding added modeling for possible Fire- documentation to Induced Initiating Events. Because it does not degrade the ability of the plant to justify the current mitigate the resulting transient, Feedwater overfeed modeling approach.
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| (MSO-N21-2ai) was included in the FPRA as an initiating This finding is event which is subsumed within the Turbine Trip initiator, sufficiently resolved ES-Al This is consistent with the treatment of initiating events in for SR ES-Al to be (NOT MET) the Internal Events model (BNP-PSA-032) and is assessed as meeting ES-A4 supported by the results of the MSO Expert Panel review. CAT 1/11/111. There is (CAT I/Il) Generically, NEI-00-01 does not list MSO-N21-2ai as no impact to the 5b FQ-A2applicable to BWR4s, noting that steam-driven feedwater application.
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| FQA2 pumps may not be a concern, and (upon review) the MSO (CAT I/Il/Ill) Expert Panel concurred.
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| Likewise consistent with the treatment of initiating events 1-9 in the Internal Events model (BNP-PSA-032), the MSO Expert Panel did not considered a plant trip to be a creditable result of a spurious HPCI operation (MSO-E41-2u). However, the possible effect of spurious HPCI operation (MSO-E41 -2u) on the ability of the plant to mitigate an otherwise initiated transient was considered.
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| In particular, during a postulated spurious HPCI operation (MSO-E41-2u), the high RPV water level signal may not
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact isolate the steam inlet valve, but Operating Experience suggests that the turbine would over speed on low quality steam and mechanically trip at some point prior to the RPV water level actually reaching the steam lines.
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| Consequently, RPV water level is not anticipated to induce a concurrent RCIC failure. However, since the available Operating Experience does not specify the RPV water level at which the steam quality is assured to cause a turbine trip, the RPV water level is identified as a source of uncertainty.
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| Documentation to justify this position has been added to Section 3.1, Attachment 3, and Attachment 8 of the
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| ________________________component selection calculation, BNP-PSA-085.
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| SY-B5 Instrumentation included in the The power supplies for the instrumentation credited for Resolution of this (CAT I/il/Ill) FPRA that affects HFES are operator actions, as identified in BNP-PSA-084, have finding added the SY-A6 listed in Calculation BNP-PSA- been added to the FPRA. The revision is documented in power supplies for the (NOT MET) 084, Revision 1, attachment 4. the component selection calculation (BNP-PSA-085) instrumentation S-9 This attachment provides a model change log. Power supplies were already included credited for operator (CAT I/l I/Ill) comprehensive list of instruments in the component selection to support other modeled actions. This finding affecting each of the modeled equipment. is sufficiently resolved ES-Cl HEPs in the PRA. for SR SY-A6 to be (CAT I/Il/Ill) assessed as meeting PRM-B9 However, the power supplies for CAT I/Il/Ill. There is (CAT I/Il/Ill) the instrumentation added to the no impact to the 5b FPRA model is not included in the application.
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| FPRA logic 1-10
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| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding ]Resolution 5b Impact Change Package BNP-0122 The BSEP trip assessment in Change Package BNP- Resolution of this includes a list of plant areas, and 0122 (Attachment 10 of BNP-PSA-080) was updated finding updated the an evaluation of a possible plant using additional insight of targets in each fire BSEP trip trip for each area. The categories compartment/zone. Targets identified for both the safe assessments using include near certainty plant trip shut down and the fire probabilistic assessment were insights of targets in (1.0), reduced likelihood trip (0.1) considered. The likelihood of a plant trip due to fire was each fire and plant trip not likely (0.01). evaluated on an area basis and considered both the compartment. There In discussions with the engineer equipment located in the area and the equipment that is no impact to the 5b who developed this list, the could be affected by traced cables that traverse the application.
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| assessment was based on area. The evaluation considered the possible effects on judgment, which including some the plant given a fire in the area but not the likelihood of a consideration for the likelihood of fire in that area or the assumed status of the system in the fire, some consideration of the the FPRA. To accommodate operator discretion, no area possible damage of a fire, and the was assigned a conditional trip probability of zero, even if equipment in the area. However, the area contained no equipment important to plant PRM-B4 the judgment did not include a operation.
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| (CAT I/Il/Ill) review of cables and equipment impacted in each area.
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| The results show that areas 1-14 impacting safety busses (which would result in a likely rapid plant shutdown), are estimated to shut down the plant 10% of the time, while impacts from spurious operation (e.g., SRV openings, MSIV closures, etc.) are not accounted for. Additionally, the base assumption of all fires causing a plant trip, loss of feedwater, loss of condenser vacuum, and MSIV closure (e.g.,
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| no cable tracing for these initiating events) is not applied.
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| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding JResolution [ 5b Impact The transient ZOI approach was 1)The turbine buikldng was re-examined for a transient fire ln the NFPA 805-based on the 75th fire versus the representing a 317 kW 98% HRR, as documented in Safety Evaluation, the 98% fire. As a result, the transient Attachment 16 of Revision 1 of BNP-PSA-086. To NRC staff found the scenarios were impacted as support the use of a lower HRR in specific areas and use of the lower HRR follows: consistent with the clarified guidance in Section G.5 of to be acceptable NUREG/CR-6850, as endorsed by the NRC in a letter because the a
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| : 1) Scenarios were rnot identified in dated June 21, 2012 (i.e., ADAMS Accession Number characterization of the areas where the cable trays were ML12171A583), an evaluation was performed as past plant-specific above 6 feet, but below the zone documented in BNP-PSA-086, Attachment 25. That transient combustible of influence for a 317 kw fire evaluation included a review of plant records for the violations was (height depends on location). performance of OFPP-01 3, Transient Fire Load provided indicating Evaluation, covering a period of two years. From those that the lower HRR is
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| : 2) Area for the ZOI was limited. records, a sample of 10 were selected for more in-depth achievable and For example, in the cable room, review, which consisted of plant walk downs, operator because an improved the area for each transient interviews, interviews with the responsible engineer, and procedure was FSS-A1 scenario was typically 3' x 3', a verification of plant wide training. In the Service Water established to provide (NOT MET) versus a longer area which may Building, where plant records repeatedly indicated confidence that all fire impact a particular cable tray. elevated combustible material, the walkdown included the loads will be limited to Again for this area, several cable inspection of the actual burnable material to determine the lower HRR.
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| 1-19 tray runs are 30' or longer, where the lower HRR to be reasonably realistic and bounding. This finding is the area assumed for a larger ZOI Where the higher HRR was determined to be more sufficiently resolved would be something like 30' x 7.' reasonably realistic and bounding for a particular fire for SR FSS-A1 to be compartment, the evaluation also considered whether the assessed as meeting
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| : 3) Areas, such as the Battery fire compartment contained equipment or targets that CAT I/Il/Ill. There is Rooms have no identified might be impacted by the higher HRR and whether further no impact to the 5b transient scenarios. evaluation would be required. Where necessary, the application.
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| evaluation identified specific fire compartments to be subjected to future administrative controls. Subsequent to the evaluation described in Attachment 25 of BNP-PSA-086, the Fire Protection Program System Health Reports covering the last three years were reviewed for violations of the transient combustible controls. These reports document and evaluate both internal plant records
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| ('e..g., Condition Reports) and NRC inspection records
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| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions (Resolution SR/F&O # jjFinding 5b Impact (e.g., by the NRC resident inspector or during triennial inspections). Condition Reports concerning events that occurred outside the Global Plant Analysis Boundary (GPAB) or during an outage were screened as not applicable to the Fire PRA. Condition Reports concerning events that occurred in the Turbine Building, where the higher HRR was used, or in Fire Compartments that were qualitatively screened in the Fire PRA, were also screened as not applicable to the use of a lower transient HRR. The evaluation also eliminated from further consideration those events where the corrective action was likely sufficient to preclude recurrence and those events that would likely not be considered a violation under Fl R-NGGC-0009. Because FlIR-NGGC-0009 includes both provisions for supporting the reduced HRRs credited in the Fire PRA and a list of specific combustible materials that have been pre-evaluated to be acceptable within those constraints, none of the violations so identified could have resulted in a transient fire exceeding the reduced HRRs credited in the Fire PRA. For the remaining Condition Reports, the evaluation considers what administrative controls are applicable under FIR-NGGC-0009 and qualitatively what effect on risk a fire might have if those controls should fail. The use (post-transition to NFPA 805) of the more proactive FIR-NGGC-0009, is credited with limiting the placement of transient combustibles and ignition sources near equipment and cables unless a specific evaluation is performed using a 317 kW 98% HRR.
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| : 2) The floor area applied for each transient scenario is based on the identified target set. The minimum applied transient foot print is 3'x3'.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact
| |
| : 3) All plant areas were assigned transient ignition source(s). If the transient ignition source did not damage any significant targets, no risk increase would be recorded from that potential fire source. This was the case for the Battery Room.
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| Transient scenarios are identified Transient and other fire impacts on sensitive electronics Resolution of this using a ZOl assuming cable were incorporated into the EPRA in Revision 5 of BNP- finding justified the damage only. No damage to PSA-080, as documented in Attachment 54. Otherwise, current modeling equipment appears to be consistent with the guidance in H.2 of NUREG/CR-6850, approach. This assumed for any area. all of the ZOIs are based on cable damage. It would be finding is sufficiently FSS-A1 very conservative to assume equipment damage based resolved for SR ESS-(NOT MET) For example, a transient fire in on the same ZOI. In most cases the equipment is Al to be assessed as the battery room was not shielded to some degree by a steel enclosure, such that meeting CAT 1/1l/Ill.
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| developed where the transient internal damage would be minimal from an external There is no impact to 1-20 damages or ignites the batteries, source. Also, exclusion zones exist to limit placement of the 5b application.
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| which is near the floor. Another unattended transient ignition sources next to MC~s /
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| example is there are no scenarios energized equipment (ref. 0FPP-014 and FIR-NGGC-located between 1CB and 1 CA, 0009).
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| where damage to both cabinets may occur.
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| HR-El No new actions were identified in Following the Peer Review a detailed HEP was Resolution of this (NOT MET) response for the Fire. developed to provide a more realistic evaluation of a finding developed an HR-E2 Discussions with BSEP remote shutdown following control room abandonment. A HEP for the control (NOT MET) operators, the ASSD procedures review of ASSD-01 and ASSD-02 identified key operator room for the will be used for shutdown given a recovery actions and related system interfaces. A proper implementation of HR-E3 fire and damage to ASSD understanding of system operation within the context of a ASSD-01 and ASSD-(CAT I,11II/ll) equipment. For example, ASSD- fire scenario was obtained during focused talk-throughs 02. This finding is HR-E4 01 will call for shutdown outside and operator interviews. The results of the HRA including sufficiently resolved (CAT I,11II/ll) of the control room in ASSD-02. the operator interviews, are documented in Attachment 10 for SR HR-El, SR
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| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact FSS-B1 For other areas, there are specific of BNP-PSA-084, Revision 2. HR-E2, and SR HRA-(CAT I/Il/Ill) ASSD procedures. ASSD-05 was Possible conservatisms associated with not modeling A2 each to be HRA-A2 reviewed for fire in Unit 1 RX'Bld other ASSD actions are not considered to be significant. assessed at meeting (NOT MET) North. This procedure includes CAT 1/11/111 and for SR specific recovery actions and HRA-D1 to be HRA-C1 manual actions, including for assessed as meeting (CAT I, II) example operation of the SRVs CAT II. There is no HRA-D1 from the RSP. Neither the control impact to the 5b (NOT MET) room evacuation actions or the application.
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| local manual actions were identified or reviewed as a part of 1 -24 the fire PRA. As a result, the FPRA results are conservative.
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| For example, the top cutset for the cable room could be recovered using a control room evacuation action.
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| The control room abandonment A detailed Human Event Probability (HEP) has been Resolution of this HEP for habitability scenarios developed for the control room abandonment scenario finding developed an uses a CCDP of 0.1 and CLERP due to habitability concerns, as documented in HEP for the control HR-G1 of 0.01, without detailed analysis Attachment 10 of BNP-PSA-084. An evaluation of the room abandonment (CAT I) or support. These values may be various key operator actions contained in the scenario. This finding FS-2 conservative or non-conservative, abandonment procedures was performed using the is sufficiently resolved (CTI) depending on the scenario CBDTM/THERP (CT, I) (including equipment damage) Calculator. Themethodology contained evaluation uses in the HRA Safe Shutdown timing for SR HR-G1 assessed to be as meeting HRA-C1 and timing. studies and feasibility analysis. The specific training and CAT I1. There is no (CAT 1,II) frequency of training was evaluated as well as a detailed impact to the 5b No detailed timing, feasibility, review of the procedure. Significant equipment failures application.
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| review of training, review of were also considered in the determination of the CCDP.
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| 1-26 procedures, or detailed task The HEPs resulting from the HRA calculator evaluations analysis was documented in the were then placed into an event tree with supporting top FPRA. logic to determine an overall HEP.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact MOO fire scenarios do not include When an open MOO fire was modeled, it is conservatively Resolution of this propagation from one MOO stack assumed that the entire MOO is failed and all targets finding justified the to another. NUREG/OR-6850 within the ZOI of the MOO are treated as failed by the fire. assumption that includes a propagation model, When the cabinet remains closed, the entire MOO is breaching with where propagation is assumed failed and the fire is assumed to remain confined to a subsequent normal FSS-A1 following a 10-15 minute delay single stack. Except that a certain probability of arcing - fire growth is possible (NOT MET) (depending on the opening). induced breaching is postulated for closed MO~s with even for closed subsequent normal fire growth. Reference FAQ 14-0009. MO~s. This finding is The BWROG methods (not sufficiently resolved 1-30 approved) includes a probability for SR FSS-A1 to be of propagate and an approach for assessed as meeting limiting the number of cabinets OAT 1/11/111. There is considered in propagation, and no impact to the 5b an approach for determining the application.
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| HRR.
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| * Location factors w(e.g., wall New walkdowns were performed for transient sources in Resolution of this effects) were not included in HRR the Turbine Buildings. The new walkdowns increased the finding accounted for calculations for transients. HRR and accounted for wall effects. These have been wall effects for incorporated into the Fire Scenario Data calculation transients and
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| * Review of the documentation (BNP-PSA-086). Use of wall effects in other areas has incorporated 211lkW shows that ceiling jet treatment been added as an uncertainty in the calculation. HRR for pumps into FSS-C1 was not performed. the FPRA. This The identification of targets from is based on a zone (OAT I)
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| * The 75th and 98th percentile influence (ZOI) determined the source HRRof using finding resolvedis for sufficiently SR ESS-HRR assigned for pumps accepted and approved methods. Where secondary fire 01 to be assessed as 1-2 (electrical fire) are from Case # 7 growth is expected, The ZOl treatment is conservatively meeting OAT II.
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| 1-2 for motors, BIN 14 (69 kW) in lieu extended to the ceiling. The treatment of ceiling jets There is no impact to of from Case # 6 for pumps, BIN would only be addressed when more detailed fire the 5b application.
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| 21 (69 and 211 kW). modeling is applied. This does not typically apply to transient sources since the overall transient analysis is based on virtual sources and does not contain the specific inputs that would be needed to justify the applicability of detailed analysis.
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| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution 5b Impact New target wakdowns were performed using 211 kW for the HRR for pumps and were incorporated into Revision 5 of BNP-PSA-080, as Attachment 44.
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| A screening value for rated The BSEP fire quantification calculation has been revised Resolution of this barrier probability of 1 E-2 was and the screening of HGL Multi Compartment Analysis finding performed the FSS-G2 applied. This may not be has been performed in accordance with NUREG/CR6850. screening of HGL (CT11/1) bounding depending on the The screening value of 0.1 was used on the exposing MCA in accordance (CTI/lIl) features of the barrier (doors, compartment to screen out compartments from the MCA with NUREG/CR6850.
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| penetrations, dampers). analysis. The results of the revised Multi-Compartment There is no impact to 1-34 Analysis is documented in the quantification calculation the 5b application.
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| BNP-PSA-080.
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| QU-B2 Truncation in the CDF and LERF The truncation approach has been changed in Rev 1 of Since the process for (NOT MET) was varied, based upon the the quantification calculation (BNP-PSA-080) in response establishing truncation QU-F2 CCDP/CLERP. For example, to this F&O. Scenarios are now run at an effective limits does not (CAT 1/l1/Ill) CCDP of 1 .0 uses a truncation of truncation of 1 E-09/yr for CDF and 1E-1 0/yr for LERF demonstrate that the U-B 1.0, while a CCDP of 1E-03 uses which is more than four orders of magnitude below the overall model results QU-B3ino E-7 vrltereutn D n LR ln oas cnegS UB a(rnaioNf1OT7 process using the ones vral (OME) h rslig D nMLRElnttTl.)ovreS UB run will continue to be FQ-B1 results in difficultly running assessed as NOT (CAT I/1l/I1l) FRANC at a very low cutoff. MET. However, the FQ-F1 very low effective (NOT MET) A review of the truncation levels truncation (i.e.,
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| was performed. Hundreds of the relative to the sequences have truncation within resulting CDF and 1-36 a factor of 100 or less of the LERF plant totals)
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| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact CCDP. Several of these provides reasonable sequences were re-run, and the assurance that no new CDFs were compared to the significant accident original CDFs. Changes in the sequence was results vary from about 5% to as inadvertently much as 25%. Many of the eliminated. This sequences affected are in the top finding is sufficiently 25 fire sequences. resolved for both SR QU-B2 and SR EQ-Fl Additionally, a large number of to assessed as scenarios are listed with zero meeting CAT I/Il/Ill.
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| COD P. When these were re-run There is no impact to with lower truncation values, the 5b application.
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| cutsets were generated. This can be important for scenarios with higher ignition frequencies.
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| LE-G2 A quantitative evaluation of LERE A quantitative evaluation of parametric uncertainty for Resolution of this (NOT MET) uncertainty was not included in both CDF and LERF was performed as documented in finding provided the LE-F3 the final results. The uncertainty EVAL EC 296040, including a State of Knowledge required quantitative (NOT MET) quantification was performed for Correlation covering fire ignition frequencies, non- evaluation of UN-A CDE results only. suppression probabilities, conditional failure probabilities, uncertainty. This (NOT MET) Assumptions and key areas of and fire bins. finding is sufficiently uncertainty did not include resolved for SR LE-EQ-E1 discussion of LERF, other than G2, SR LE-F3, SR (NOT MET) the use of a simplified LERF UNO-Al, SR EQ-El, EQ-Fl value for control room and SR FQ-F1 to be (NOT MET) abandonment. assessed as meeting CAT 1/11/111. There is no impact to the 5b 1-38 ____________________________________application.
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| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The BSEP EPRA roadmap In the process at BSEP, Fire Protection/NSCA develops Resolution of this CS-Al indicates that the methodology to and maintains the cable selection and circuit analysis finding justified the (CAT I/ll/Ill) identify additional cables uses data. These data are then referenced as inputs to the current methods. This CS-A3 same process for PRA circuit Component Selection and Quantification FPRA finding is sufficiently (CAT 1/11/111) analysis as for the deterministic calculations. This process and associated results are resolved for SR CS-CS-Cl Safe Shutdown circuit analysis. easily reviewable, has been peer reviewed multiple times C1 to be assessed as (NOT MET) Reference FIR-NGGC-01 01. for our other sites and found to be acceptable. There is meeting CAT 1/11/111.
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| However, there is no separate no requirement to have a separate PRA notebook. There is no impact to notebook for Fire PRA Cable the 5b application 2 Selection to discuss the processes, inputs and results.
| |
| The QLS screening criteria may Revision 2 of BNP-PSA-083 removed FC261 Resolution of this not have been applied (DUCTBANK) from qualitative screening and retained it finding removed appropriately. BNP-PSA-083 Rev, for quantitative analysis. FC261 (DUCKBANK) 1 Section 3.3 documented the from qualitative QL-1 screening criteria as in Raceway target information, cable loadings and floor screening and (CSAT/lll) NUREGICR-6850. Alternate areas for the manholes in the DUCTBANK were collected, retained it for (CTII/l) screening criteria was used to MOS factors were assigned, and transient ignition quantitative analysis.
| |
| QLS-.A2 screen several analysis units. frequencies were determined. There is no impact to (CAT I/Il/Ill) These criteria were based on the the 5b application.
| |
| QLS-A3 judged low risk significance of the With the exception of FC261 (DUCTBAN K), no physical (CAT I/Il/Ill) unit in question. analysis unit was qualitatively screened based on the use QLS-A4 of alternate screening criteria. However, FC295 and (CAT I/Il/Ill) Moreover, FC261 (DUCTBANK) FC345 (i.e., DrywelliTorus, for Unit 1 and Unit 2, QLS-B3 was screened out based on no respectively) were not retained for quantitative analysis (CAT I/Il/lll) equipment while the QLS because no ignition frequency was assigned to the screening criteria need to rule out Drywell/Torus based on the Technical Specifications both equipment and cables. requirements for an inert atmosphere during power 2-3DUCTBANK will contain a large operations. This treatment is consistent with both the Fire number of cables and low risk Hazard Analysis in the (U)FSAR and the Safe Shutdown contribution is not expected. Analysis in which no fire is postulated in or analyzed for the Drywell.
| |
| BSEP team responded as follows:
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # jFinding Resolution j 5b Impact
| |
| 'FC261 (DUCTBANK) is not a typical fire compartment (i.e., '. ..a well-defined enclosed room...').
| |
| As described in Attachment 3, FC261 (DUCTBANK) is a network of underground conduit in pre-cast concrete cable trenches.
| |
| Rather subsuming FC261 (DUCTBANK) into FC263 (with certain 'yard' locations), FC261 (DUCTBANK) was separately identified during plant partitioning to promote clarity in communication with legacy plant fire protection programs. Because of its design, no transient fire was postulated for FC261 (DUCTBANK). As described in Attachment 3, no equipment is located in the FC261 (DUCTBANK). And as stated in Section 3.4.2, all cables at BSEP are considered qualified self extinguishing and non-propagating.
| |
| With no creditable ignition source, there is no fire risk. Therefore, it was considered appropriate to qualitatively screen FC261 (DUCTBANK) consistent with the stated intent of the general task objective described in Section 4.3.1 of NUREG/CR-6850.'
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact However, it is not expected to have absolute zero fire ignition frequencies in the ductbanks since these enclosed areas could be open for maintenance during outages and transients could be left there unnoticed, even the transient materials from plant startup. On the other hand, 100%
| |
| qualified self extinguishing and non-propagating cables may not be realistic. Cables used for lighting and other not modeled system functions may exist in the ductbanks, which may not be
| |
| _____________ qualified.
| |
| SY-C1i The system notebooks (i.e., A discussion of fire impacts was added as Section 3.7 of Resolution of this (NOT MET) calculation BNP-PSA-062) have BNP-PSA-062, Revision 11, and included fire-related finding added a SY-A2 been updated for MOR1 1 from changes (e.g., disposition of MSOs) to the FPRA and discussion of fire (NOT MET) which the FPRA was sources of model uncertainty and related assumptions. impacts and fire S-2 subsequently developed. The related Changes to the SY2 system notebooks (i.e., system notebooks.
| |
| (NOT MET) calculation BNP-PSA-062) will be This finding is SY-A3 further updated to incorporate sufficiently resolved for (NOT MET) fire-specific changes to the SR SY-C1, SR SY-A2, SY-A4 model. SR SY-C2, SR SY-A3, (CAT I) and SR SY-A6, to be SY-A6 However, the system analysis assessed as meeting (NOT MET) supporting requirements included CAT 1/11111l and for SR PR-9 in SY-A2, A3, A4, A6, C1 and C2 SY-A4 to be assessed (CARMll/ll have been determined to be not as meeting CAT IlI/ll.
| |
| (CT11/1) met with the current There is no impact to PRM-C1 documentation, which was
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution j 5b Impact (CAT 1/11/111) typically performed by updating the 5b application.
| |
| the system notebooks to reflect all fire-related changes.
| |
| 2-8 An example of information not included from SY-A2 includes:
| |
| COLLECT pertinent information to ensure that the systems analysis appropriately reflects the as-built and as-operated systems.
| |
| Examples of such information include system P&IDs, one-line diagrams, instrumentation and control drawings, spatial layout drawings, system operating procedures, abnormal operating procedures, emergency procedures, success criteria calculations, the final or updated SAR, technical specifications, training information, system descriptions and related design documents, actual system operating experience, and interviews with system engineers and operators.
| |
| See other referenced SRs for other information not included in the FPRA documentation.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The system notebooks A discussion of fire impacts was added as Section 3.7 of Resolution of this (calculation BNP-PSA-062) have BNP-PSA-062, Revision 11, and included fire-related finding added a been updated for MOR1 1 from changes (e.g., disposition of MSOs) to the FPRA and discussion of fire SY-C3 which the FPRA was sources of model uncertainty and related assumptions. impacts and fire (NOT MET) subsequently developed. The related changes to the PRM-C1 system notebooks (calculation system notebooks.
| |
| (CAT I/Il/Ill) BNP-PSA-062) will be further This finding is updated to incorporate fire- sufficiently resolved for specific changes to the model. SY-C3 to be assessed 2-9 However, the sources of model as meeting CAT I/Il/Ill.
| |
| uncertainty and related There is no impact to assumptions are not documented. the 5b application.
| |
| PRA items that were assumed The requested sensitivity is on items considered always The current failed for the component selection failed in the Fire PRA. This treatment represents a assumption is QU-E4 are listed in BNP-PSA-085 Rev. 1 conservatism in the calculated Fire CDF. Of these, the identified as a (CAT 1/11/111) Section 4 and Table 4. largest effect is likely the assumption of loss of feedwater conservatism in the UNC-A1 This treatment is similar to for each scenario. calculated Fire CDF.
| |
| (NOT MET) treatment of unknown locations The finding is PRM-B10 for equipment that do not have sufficiently resolved (CAT I/li/Ill) cable-routing completed, for SR UNC-A1 to be Sensitivity studies should be assessed as meeting performed to investigate the risk CAT I/Il/Ill. There is 2-10 importance of these failed no impact to the 5b systems/functions. application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact BNP-PSA-080 Section 4.5.3, Section 3.2.3.3 of BNP-PSA-083 documents Resolution of this Non-Suppression Probability, consideration of the applicability of using generic non- finding justified the documents the methods used for suppression data based on an outlier review of plant fire current methods. This calculation of non-suppression bridge experience. To support the FPRA Peer Review a finding is sufficiently probabilities. Generic NSP and review of the then-current one year period (i.e., 2011) of resolved for SR FSS-unavailability are applied from System Health Report information was performed and D7 to be assessed as NUREG/CR-6850. No outlier indicated no "outlier behavior" for the Fire Detection and meeting CAT I1.
| |
| review is performed, and no plant Suppression Systems. This focus on a 1-year period There is no impact to specific data are used to update provided an overview of the most current system the 5b application.
| |
| the unavailabilities. performance, measured against specific parameter/attributes, and helped to confirm the effectiveness of preventative and corrective maintenance.
| |
| FSS-D7 During the NFPA 805 RAI response, the most recent (CAT I) three years of System Health Report information were also reviewed and found to show sustained acceptable performance levels, again with no "outlier behavior" noted 2-14 for the Fire Detection and Suppression Systems.
| |
| Currently, system performance is monitored and maintained at a high level as part of the System Health Reporting and System Notebook processes. Outlier behavior with respect to system availability would be evident to the system engineer and plant management through the health data (available for the previous 12 months), which indicates overall Excellent (Green) performance. Post-transition, the assessment of system performance is part of the NEPA 805 Monitoring Program, as described in procedure FIR-NGGC-0130.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SRIF&O # [Finding Resolution J 5b Impact The note of SR FSS-D8-1 states: Accommodation of area specific features and scenarios is Resolution of this Fire detection or suppression assured for fire suppression and detection system finding reinforces the system effectiveness depends on, through correct application of fire protection design correct application of at a minimum, the following: 1) standards such as NFPA 13- Standard for the Installation fire protection design system design complies with of Sprinkler Systems, and NFPA 72- National Fire Alarm standards which applicable codes and standards, Code. In each case careful selection of occupancy ensures that physical and current fire protection classification and hazard identification is applied. This features and the fire engineering practice, 2) the time ensures that physical features and the fire sources sources contained in available to suppress the fire prior contained in a given area are properly protected to a given area are to target damage, 3) specific achieve the desired performance results. Ceiling properly protected to features of physical analysis unit configurations, blockage of agent application by design achieve the desire and fire scenario under analysis features and adequate coverage for the hazards present performance results.
| |
| (e.g., pocketing effects, are a direct function of code compliance. Code There is no impact to blockages that might impact compliance is further assured by detailed evaluation in the 5b application.
| |
| plume behaviors or the "visibility" the NFPA 805 Transition report Table B-i, th rough the FSS-D8 of the fire to detection and use of or reference to Code Compliance calculations such (I/Il/Ill) suppression systems, and as 0FP-1038, Rev. 1, Code Compliance Evaluation NFPA suppression system coverage), 13 (Reactor Building), 1976 and 1983 Ed. or 0FP-1 031, and 4) suitability of the installed Rev. 0, Code Compliance Evaluation NFPA 72E, 1984 2-15 system given the nature of the fire Ed. for BSEP.
| |
| source being analyzed.
| |
| In light of BSEP fire scenarios, above item 1 should be considered met although not evident in documentation. Timing (item 2) is considered in detailed NSP calculations were carried out in the spreadsheet files BN P_EVAL_Ul1_CDF.xls, BNP_EVAL_UI_LERF.xls, BNP_EVAL_U2_CDF.xls, and BNP_EVALU2_LERF.xls.
| |
| .1____________________________________________________________ .1________________________
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact However, above items 3 & 4 were not addressed.
| |
| BNP-PSA-086 R0 Section 10.0 Section 4.13 of Calculation BNP-PSA-095 evaluated the Resolution of this states that fires resulting in treatment of smoke damage, as discussed in Appendix T finding evaluated the significant smoke production of NUREG/CR-6850, and the possible effects on the potential for smoke could cause additional damage quantification results in the BSEP FPRA. Appendix T of damage to EPRA beyond the heat based zone of NUREG/CR-6850 limits the equipment vulnerable to short equipment on a influence target sets collected. term smoke damage to medium and high voltage qualitative basis and FSS-D9 However, targets that are switching or transmission equipment, and lower voltage incorporated the (CAT I) susceptible to smoke damage instrumentation and control devices, results of the have not been identified and are The BSEP FPRA currently accounts for smoke damage assessment into the currently not evaluated in this consistent with the guidance in Appendix T of definition of fire calculation. Therefore, this SR is NUREG/CR-6850 by failing the entire electrical bus or scenario target sets.
| |
| 2-6 considered not met for CC-lII. panel where the fire is postulated. This accounts for any This finding is smoke damage generated inside the panel. sufficiently resolved for SR FSS-D9 to be assessed as meeting CAT li/I/ll. There is no impact to the 5b application.
| |
| A review of FRANC model files The FPRA database query qSourceBE 2a(source) Resolution of this showed that some HGL scenarios which identifies the failed components for the individual finding modified the FQ-A1 (whole room burnout) have less scenarios was modified to use the same FSSPMD existing mapping (i/Il/Ill) affected components than some mapping table (TRoutingFireZone) that is used for tables for use in both individual scenarios in the same generating the HGL component failures (Reference BNP- individual scenarios fire compartment (FC) modeling a PSA-080 Rev 1). and HGL scenarios.
| |
| 2-19 single ignition source and targets There is no impact to in its ZOI. the 5b application.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding [Resolution 5b Impact For example, in Unit 1 CDF FRANC model, FC212 scenario BHGL has 64 affected components while scenarios FC212_4612 B75 and B98 have 112 affected components.
| |
| On the other hand, some other scenarios have significantly more affected components in HGL scenarios than individual scenarios in the same Fo.
| |
| Discussion with BSEP PRA team indicated that different mapping tables have been used for HGL scenarios and individual ignition source scenarios. Conservatism may exist in the generation of mapping tables for individual scenarios. However, non-conservatism could exist for the HGL scenarios if the different mapping tables do not cover all the cables / equipment that are affected by the fire-induced failures.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact FC238_5010 and similar fire Metal water spray shields are provided over several fire Resolution of this scenarios are expected to have initiators in the Diesel Generator Basement. Specifically finding verified plant significant SBO contributions. A these metal shields are installed over the EDG Excitation design, addressed review of mapping table and Voltage Source PT & Reactor and EDG Excitation potential failures, and excluded events and altered Current XFMR Phase A, B, & C as documented in enhanced the event tables did not show the drawing F-1319. The spray shields are BSEP plant documentation failure of DG Breaker spurious configuration and are maintained via controlled drawings. concerning the metal failures excluded, which is also The shields are designed to prevent water spray water spray shields.
| |
| evident in FRANC affect impingement onto the transformers described above and There is no impact to components. However, it is noted per controlled drawing examination and plant walkdown the Sb application.
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| that the IGN is set to 0. they also provide a non-combustible barrier to the development and passage of a damaging fire plume BSEP PRA team responded that: above these transformers. Based on walkdowns the
| |
| 'During review of cutsets following construction of these shields is sufficient to prevent direct preliminary quantification, several passage of a damaging fire plume to targets located EQ-Al scenarios were identified as directly above the protected transformers.
| |
| .(I/il/Ill) significant contributors to plant risk. Review of these scenarios The primary concern with a fire in the subject identified significant transformers (i.e., sources 5010, 5011, 5012, 5013 5014, 2-20 conservatisms in the initial data 5015, 5016 and 5017) is development of a fire plume that inputs that were causing would impact cable trays routed above the spray shield.
| |
| unrealistic risk results. As part of The design of the spray shield is such that the plume this review, it was identified that would be forced to follow a circuitous path prior to the fire size for sources 5010 impingement on the target cable tray. The worst case fire through 5017 were initially expected to develop in the fire initiators would be a 69kW characterized as 211 kW fires fire based on the 98% HRR for dry-type transformers, when a more detailed Ref. NUREG/CR-6850.
| |
| examination of the equipment showed that the sources should All of the cables located in the Diesel Generator be characterized as 69 kW fires. Basement are IEEE 383 qualified; therefore their damage In addition, a shield above the temperature is 625°F and damaging heat flux is 11 kW/m 2 .
| |
| sources was identified. The target cable trays are located above the EDG Consideration of either of these transformer spray shields therefore damaging two facts would result in temperatures must be exceeded at the spillage points of
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # jFinding ]Resolution J 5b Impact consequences for the fires that the spray shield to be deemed capable of damaging the would be much less severe than cable trays or a damaging radiant heat flux radiated from the initial walkdown information the spray shield. Based on review of the spray shield indicated. This information is design and plant walkdown of the initiator/target documented in change package configurations it is judged that the spray shields installed BNP-0182 and BNP-0176 (See above these transformers will prevent thermal damage to the BNP-0176 change package the target cable trays, thus damage resulting from a directory of BNP-PSA-080 calc transformer fire need not be postulated.
| |
| for pictures of these sources).
| |
| Because the quantification Continued maintenance of these spray shields is ensured process was nearly complete, by plant documentation and credit for these spray shields explicitly incorporating the as a radiant/plume shield for raceways located above the information from the change EDG transformers is documented in the fire PRA packages into all input calculation.
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| calculations would have resulted in a significant administrative burden to revise the calculations.
| |
| Therefore, to simulate the correct effects within the quantification calculation, the scenarios were assumed to be equivalent to the first target tray having a solid bottom as per BNP- 0176 and the scenario event frequency was set to zero for scenarios FC238_5010 B75 and FC238_5010 B9B. It is assumed that cable trays with solid bottoms will prevent damage to cables for ignition sources with HRR 69 KW or less based on the discussion provided in section Q.2.2 of NUREG/CR-6850.'
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding _Resolution j 5b Impact A test fire model run with FDS was also constructed to demonstrate the adequacy of above engineering judgment. As a result, the technical basis supporting the treatment of the identified scenarios is considered acceptable. However, the following issues should be addressed:
| |
| : 1. The documentation in BNP-0176 should be enhanced to include engineering judgment as discussed above instead of a simple assumption that metal cover above the cabinet is sufficient in preventing fire damage to targets above the cover.
| |
| : 2. The BSEP team stated that the metal cover is part of the design basis. This fact should be verified and documented in fire PRA.
| |
| : 3. The potential failure of the metal cover should be addressed.
| |
| May need to credit the surveillance / inspection /
| |
| maintenance program to ensure the integrity of this metal cover.
| |
| : 4. Perform sensitivity study or include the failure probability of this metal cover to generate risk insights associated with the assumption associated with this
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions S R/F&O # Finding Resolution 5b Impact metal cover.
| |
| : 5. Revise the fire PRA model to not set ignition frequency to 0 but remove the impacted targets.
| |
| BNP-PSA-080 Section 4.3.4, Fire The listed non-instrument spurious cable failures were Resolution of this Induced Spurious Event analyzed, and probabilities were included in the Fire PRA. finding analyzed the Probabilities, document the Conditional failure probabilities were assigned to the most listed non-instrument methods used for conditional risk significant contributors, causing them to become less spurious cable failures failure probabilities for fire- risk significant and allowing these less risk significant and the probabilities induced circuit failures, contributors to appear relatively more risk significant. were included in the More could have been done, but the iterative process Fire PRA. This Circuit Analysis was performed in stopped when satisfactory results were obtained, finding is sufficiently CF-Al change package BNP-0137 to resolved for SR CF-(CAT I) determine the probability of a In many of the identified cases, failures are in Al to be assessed as CF-B1 spurious operation for various instrumentation, and probability analysis methods are not meeting CAT IlI/ll.
| |
| (CTIl/l) cables, available, and no testing has been done to determine the There is no impact to (CT /I/Il)failure probabilities. Division of failure mode based on the 5b application.
| |
| Risk significant contributors were conditional probability analysis would only serve to add not identified (quantification was additional uncertainty to the failures.
| |
| 2-22 complete later in the process) and utilized thus cannot met the The current analysis is conservative in that for cases capability category CC-Il. where specific conditional probabilities have not been developed, failure or spurious operation is give a For example, the Unit 1 CDF probability of 1.0.
| |
| importance results include the following spurious events for which conditional probabilities have not been developed:
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding jResolution f 5b Impact HPC1 PPS-SA-N12A_TPRESSURE SWITCH E41 -N012A SPURIOUSLY ACTUATES HPC1 PPS-SA-N12CTPRESSURE SWITCH E41-N012C SPURIOUSLY ACTUATES RCI 1TME-HI-N021B_TTEMPERATURE ELEMENT E51 -TE-N021 B SPURIOUS OPERATION RCI 1TME-HI-N022B_TTEMPERATURE ELEMENT E51-TE-N022B SPURIOUS OPERATION RCI 1PPS-SA-N012ATPRESSURE SWITCH E51-N012A SPURIOUS OPERATION RCI 1 PPS-SA-N012C_TPRESSURE SWITCH E51-N012C SPURIOUS OPERATION HPC1 PPS-SA-N 12B_TPRESSURE SWITCH E41-N012B SPURIOUSLY ACTUATES HPC1 PPS-SA-N12D_TPRESSURE SWITCH E41 -NO12D SPURIOUSLY ACTUATES SRVlISRV-CO-F01 3G TNON-
| |
| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions SRIF&O # ]Finding -
| |
| Resolution 5b impact ADS SAFETY RELIEF VALVE B21-FO13G SPURIOUSLY OPENS RHR1 MDP-SA-COO2CTRHR PUMP E11-COO2C SPURIOUS START DUE TO FIRE RCI1 PPS-SA-NO12BTPRESSURE SWITCH E51-NO12B SPURIOUS OPERATION RCI 1PPS-SA-NO12DTPRESSURE SWITCH E51-NO12D SPURIOUS OPERATION HPC1 PPS-SA-N17ATPRESSURE SWITCH E41-NO17A SPURIOUS OPERATION HPC1 PPS-SA-N17BTPRESSURE SWITCH E41-NO17B SPURIOUS OPERATION SwS1 PPS-SAP129LTPRESSURE SWITCH PS129 SPURIOUS OPERATION FAILS LOW ISOLATES HEADER Note that if the instrument spurious operations above are not caused by a hot short, detailed circuit analysis is likely not needed. However, the valve and pump spurious operation
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact would likely benefit from additional analysis.
| |
| BNP-PSA-080, Attachment 17, The Seismic-Fire Interaction Analysis report has been Resolution of this Section 5 documents the failure updated to address this F&O. Sections 5.3.3, 6.4.3, and 9 finding updated the or spurious operation of detection have been updated to address updates since the IPEEE. Seismic-Fire and suppression systems. Interaction Analysis Flooding, habitability and life A list of suppression systems that were modified from dry- report. There is no safety concerns are also pipe to wet-pipe systems was determined from DBD-62. impact to the 5b addressed, but only through Using ESR 94-00345, it was confirmed that the previous application.
| |
| reference to the IPEEE. However, flooding analysis conducted for the plant remained valid no update to this evaluation is for these suppression systems. Therefore, the SF-A2 provided. During the walkdown, it modification of these systems did not introduce any new (CAT I/Il/Ill) was noted that some changes in flooding concerns, and the conclusions from the IPEEE SF-A3 the fire suppression system had evaluation remain valid.
| |
| (CAT I/Il/Ill) recently occurred, including changing some systems from dry The Seismic-Fire Interaction Analysis report has been to wet-pipe systems. updated to address this F&O. Sections 5.2, 5.4, 8, and 9 3-4 have been updated to address common fire pump suction The following have also not been piping.
| |
| specifically addressed:
| |
| Discuss seismic vulnerability of The piping between the diesel and motor driven fire any common fire pump suction pumps is not seismically qualified. Based on drawing piping. A common suction for review and relevant site documents, a single break in the both the electric and diesel fire suction piping from the Fire Protection Water Tank or the pump is provided from the DWT would not result in the loss of both fire pumps due to 300,000 gallon storage tank. the presence of isolation valves. If multiple breaks were to
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution Sb Impact Failure of this line can result in occur due to a seismic event, water supply to both fire failure of both fire pumps. pumps could be compromised. DWT suction piping is not considered vulnerable as it is routed underground in some areas.
| |
| BNP-PSA-080, Attachment 17, The Seismic-Fire Interaction Analysis report has been Resolution of this Section 5.2 discusses common updated to address this F&O. Sections 5.2 and 9 have finding updated the cause suppression failures of the been updated. Seismic-Fire fire water system. The common Interaction Analysis cause failure of gaseous The Unit 1 and 2 HPCI fire compartments each contain an report. There is no suppression system (002 and automatic 002 suppression system. Each system is impact to the Sb Halon) is not discussed. supplied by two banks of C02 supply tanks, designated application.
| |
| the main and reserve banks. These supply tanks are No discussion is provided in located outside the Reactor Building that they serve. Unit SF-A3 regards to establishing redundant 1 HPCI Fire Compartment FC-RB1-2 is served by the SR3 supply of fire water or gaseous main and reserve banks in Fire Compartment HCB1, and (CAT I/Il/Ill) agent supply. Unit 2 HPCI Fire Compartment FC-RB2-2 is served by the main and reserve banks in Fire Compartment HCB2.
| |
| 36Plant procedures should Each set of main and reserve banks serves only the 3-6 specifically address availability of automatic suppression system for the adjacent Reactor redundant fire water and gaseous Building.
| |
| agent supply in case of loss of the main supply of fire water or Based on the close proximity of the main and reserve normal gaseous agent supply. banks for each system, and their location in a non-seismically qualified fire compartment, a seismic event could damage both the main and reserve supply banks and cause the 002 system they supply to become inoperative. However, because the supply for FC-RB1-2
| |
| ____________and FC-RB2-2 are separated by a large, open distance,
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding ]Resolution 5b Impact there is no common cause failure that could result in the loss of supply for both automatic 002 suppression systems.
| |
| The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-O1). The Halon suppressant supply for the system in this fire compartment is local, and so common cause failure is not a concern.
| |
| Discussion of the availability and use of alternate water supply was increased in the report. These alternate supply sources include the DWT and Intake Canal, while the alternate pressure source if both fire pumps are unavailable is an external pump truck. If the fire pumps are unavailable, water supply and pressure can be maintained in the fire suppression ring by external pumper truck through yard hydrants.
| |
| Each carbon dioxide system for the Unit 1 and 2 HPCI fire compartments contain a main and reserve supply bank, but no other redundant supply was found for these systems. The Unit 1 and 2 systems do not share a common supply and cannot be cross-tied.
| |
| The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so redundant supply due to common cause failure was not examined.
| |
| Plant procedure 0OP-41 includes procedures used to
| |
| _____________align the fire protection system to alternate water supplies
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&0 # Finding Resolution 5b Impact and an alternate pressure supply.
| |
| There is a selector switch for each 002 system to select between main and reserve banks, but no procedure was found for the use of this selector switch. The operation of this selector switch should be included in a procedure to allow for transfer from the main to reserve bank (or vice-versa) in the event the selected supply bank becomes unavailable. No physical cross-tie or procedure to align one unit's HPCI C02 system to the 002 supply of the other unit was found in this analysis.
| |
| The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so procedures to align a redundant supply due to common cause failure were not examined.
| |
| Justification for partitioning Additional justification/clarification was added to BNP- Resolution of this elements that either lack a fire PSA-083 for the partitioning elements that lack a fire finding added resistance rating or have been rating, especially with regard to the presence of justification for PP-B1 omitted need to be provided for intervening combustibles for open partitioning elements, partitioning elements (CAT I/ll/Ill) the following fire compartments that lack a fire rating.
| |
| PP-B2 (examples only): This finding is (NTMT oFC207 - The east wall has an sufficiently resolved open doorway to FC206 which is for SR PP-B2 to be PP-C3 not justified assessed as meeting (CAT I/Il/Ill) oFC210O/FC211- -fire rated seals CAT IlIll1. There is no that cannot be maintained as fire impact to the 5b barriers application.
| |
| 3-8 *FC238 (DG-1) - This compartment also interfaces with FC244, 245. No justification of partitioning.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding JResolution [ 5b Impact
| |
| * Generic - Block walls are rated for 2 hours per 3.2.2.2, however, the walls column identifies them as 3 hours in most cases. Some cases no rating is provided.
| |
| * FC252 - No justification for unrated block wall - south.
| |
| *FC269, 270, 271, 272 - No justification for open grating and stairwell. The only discussion is that openings are beneficial in preventing HGL. If partitioning is not an issue, then it could be combined as one area.
| |
| Transients or fixed combustible ignition sources and intervening combustibles close to the opening may result in damaging plume temperatures beyond the compartment and/or affect OMAs and fire response.
| |
| * FC274, 275 - compartment above separated by concrete ceiling and open chase. No justifications for open pipe chase, except that it aids in preventing HGL.
| |
| * FC278, 279, 284, 285 - Open stairwell, electrical chase and pipe chase are not justified.
| |
| * FC270 is spatially separated from FC269 by the mezzanine space above HPCI room. No justification has been provided, I _________________________________________________________ I _______________________
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution j 5b Impact i.e. distance, intervening combustibles, combustible free zones, etc.
| |
| Severity factor calculations are Severity factors were applied to every scenario, based on Resolution of this based on generic data per ignition the approved calculation (NED-M/MECH-1 006). finding justified the source and the distance to the However, most had a Severity Factor of 1.0 because the current process. This nearest target (BNP-PSA-080). closest target was within the ZOI of the lowest HRR Bin. finding is sufficiently FSS-C1 Review of the BPNFPRA These distances are based on well documented walk- resolved for both SR (CAT I) database (and associated down results. Sources were typically evaluated for at FSS-C1 and SR FSS-FSS-C4 BNP_EVAL spreadsheet) shows least two HRRs based on the 75% and the 98% 04 to be assessed as (OAT I) that the distance from the ignition percentile fires. meeting CAT I1.
| |
| source to the nearest target is 0 There is no impact to inches for 3779 of the 4907 the 5b application.
| |
| 3-12 sources (including transients).
| |
| Other target distances are mostly few inches from the source.
| |
| Resulting SF is 1.0 for almost all scenarios.
| |
| Unreviewed Analysis Method This analysis method was piloted at HNP and is generally Resolution of this described in FAQ 14-0009, which was being developed finding justified the 3S-Al concurrent with NRC review of the BNP NFPA 805 LAR. current process and The BSEP FPRA calculates SS-A using: In Section 3.4.2.2 of the associated Safety Evaluation, the promulgated the jt Met) 1) A severity factor 0.1, where NRC found the 0.1 cabinet breaching factor to be method to the rest of 90% of the fires are contained acceptable for well-sealed MO~s at BNP because it is the industry through 4-1 within the MOO consistent with available operating experience and is the FAQ process.
| |
| : 2) HRR severity factors are systematically applied to a representative physical This finding is treated independently, similar to configuration. sufficiently resolved
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact other cabinets, for SR ESS-A1 to be In the final version, FAQ 14-0009 conservatively uses a assessed as meeting breaching factor of 0.23. This relatively minor difference CAT 1/11/111. There is does not invalidate the acceptability of the existing no impact to the 5b approach but is considered new data which would be application.
| |
| evaluated for incorporation into the FPRA as part of the normal model maintenance process.
| |
| BSEP cable routing information is All of the cables routed for FSSPMD contain the terminal Resolution of this contained in the BSEP FSSPMD information. In fact, the cable naming includes the finding confirmed that database. This data base termination information. There are some instances where terminal information is contains cable routing information these data have not been repeated in the FROM/TO included for all cables CS-Al10 for the selected cables and fields, however this field not required for BSEP. These in FSSPMD. There is (CAT Il) includes routing information for fields exist because in other plants the termination no impact to the 5b CS-C2 the analysis unit and raceway information must be entered specifically. application.
| |
| (CAT I/llI/Ill) information for the subject cables.
| |
| The database includes treatment of cable terminal end locations for 4-.5 most cables contained in the database. However, several cables were found with no terminal data included.
| |
| No new thermal hydraulic New engineering calculations, thermal hydraulic analysis, Resolution of this analysis was used in the and simulator runs were performed to confirm the finding performed SC-B1 construction of the fire PRA, success criteria previously established by engineering additional analysis to (NOT MET) however, there are several judgment. There was no change to success criteria confirm the success instances where engineering previously modeled that were based on engineering criteria previously PRM-B7 judgment was used to justify no judgment. The component selection calculation (BNP- established by (CAT I/llI/Ill) changes are required in the PSA-085) has been updated to reference the specific engineering judgment.
| |
| existing success criteria, calculations used in determining the success criteria. This finding is sufficiently resolved 4-8 There are several instances for SR SC-B 1 to be where thermal hydraulic analysis assessed as meeting could have been used to replace CAT II. There is no
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact engineering judgment in the impact to the 5b justification or no justification was application.
| |
| found for use of existing success criteria in the internal events criteria: 1) no evaluation of the affects on the thermal hydraulic calculation and or timing was found for MSO C1 1-2e (RPV coolant drain through the SDVvent and drain) for loss of 138,000 gal of suppression pool inventory on accident progression. 2)T23-4U (Spurious opening of torus vent and purge valves) no thermal hydraulic evaluation of long term affects of short term containment failure on long term containment over pressure. C71-lA (ATWS) -
| |
| Justification states that hot shorts may last for up to 11.3 minutes, this may have a significant impact on the thermal hydraulic analysis, this needs to be considered if this timing is used in the justification for exclusion of the MSO.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding IResolution 5b impact In general analytical methods F&O 4-9 seems to confuse expert judgment with Resolution of this were not used in the limited conservative decision making, further confounding the finding justifies the changes for success criteria for issue by labeling the later as engineering judgment and use of expert the fire PRA. All of the analysis suggesting that CATiI prohibits its use. Expert judgment judgment as opposed reviewed has some type of is defined in the standard as "information provided by a to analytical methods.
| |
| engineering judgment included in technical expert, in the expert's area of expertise, based This finding is the justification. on opinion, or on an interpretation based on reasoning sufficiently resolved MSO P41-5e is an example for a that includes evaluations of theories, models, or for SR SC-B2 to be change in the success criteria of experiments." This differs markedly with the example assessed as meeting a credited system which includes cited, MSO P41-5e, in that the MSO involves a limited CAT IlIll1. There is no engineering judgments and or number of possible outcomes (i.e., either the flow impact to the 5b assumptions for the justification. diversion fails the NSW pump or it does not). Assuming application.
| |
| that the NSW pump fails is certainly the more Case # P41 -5e
| |
| | |
| == Description:==
| |
| conservative decision. Citing hard data (e.g., pump SC-B2 Spurious operation (open) of both design capacity or operating flow rate) for the expected (CAT I)
| |
| RHR service water isolation performance of specific equipment as the basis for PRM-B7 (crosstie) valves in a loop may making a conservative decision should certainly not (CAT I/Il/Ill) result in diversion of service water cause the resultant stated assumption to be treated with flow from the RHR heat the same level of scrutiny as the, presumably much exchangers. softer, information based on an opinion formed from the 4-9 evaluation of a theory, model, or experiment.
| |
| PRA Disposition: 'Each nuclear service water pump has an 8,000 gpm design capacity. Each RHR SW heat exchanger has a design flowrate of 8,000 gpm. The RBCCW system is adjusted for a 7,200 gpm flow rate. The RBCCW system only automatically isolates on a LOCA or LOOP signal. Since LOCA's are not considered in a fire PRA, it is assumed that one nuclear service water pump is needed
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SRIF&O # Finding Resolution 5b Impact and aiigned to RBCCW at the time of the fire. It the second NSW pump automatically starts (including discharge valve opening) on low NSW header pressure, a spurious opening of one RHR HX path will be mitigated. If two or more RHR HX paths spuriously open, it will be assumed that both NSW pumps will fail due to run-out. Otherwise, if the standby NSW pump does not start of its discharge valve does not open, only one RHR HX path needs to be spuriously opened to fail the operating NSW pump. In this case the standby NSW pump will also be failed (due to the assumed valve or pump failure).
| |
| The following combinations model this MSO (and SO R43-5j).'
| |
| The BSEP approach of fire The Fire Scenario Data calculation BNP-PSA-086 has Resolution of this FSS-A4 scenario development was to been updated to address this F&O. Section 9.5.2 has finding updated the (NOT MET) evaluate all identified fire sources been updated to include fire propagation. Fire Scenario Data ESS-D1 1 individually. These fire scenarios calculation to include (CAT I/Il/Ill) included the specific cable tray, The database was updated by adding several queries that fire propagation. This FSS-G1 component, and conduit targets create tables which determine the secondary initiator finding is sufficiently (CAT I/I I/lll) for each credible source. within the most limiting ZOI. All other targets that are resolved for SR ESS-However, review of the located above the secondary initiator (larger DIST_V A4 to be assessed as information determined that the value) are then included to be in the same ZOI as the meeting CAT 1/11I/Ill.
| |
| 4-1 1 identified targets included were limiting secondary initiator. This is done by setting the There is no impact to only those within the zone of fields [69 kW], [143 kW], [211 kW], [317 kW], and [702 the 5b application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact influences of the initial source. kW] for all targets vertically above the limiting secondary No additional targets were initiator to match the same fields of the limiting secondary included that were in the zone of initiator in the table [Z Source-Target].
| |
| influence for fire growth scenarios intervening combustibles such as cable trays were in the original zone of influence.
| |
| The assessment used to quantify While conservative, CAT II can be MET if the fire risk is Resolution of this the fire risk for the unscreen bounded. finding justified the analysis compartment used fairly current approach.
| |
| conservative approaches, such Some type of fire modeling was performed for all ignition This finding is as consideration of only the 75% sources and the development of the FPRA included the sufficiently resolved and 98% fires, inclusion of selection and application of either computational or for SR FSS-D3 to be suppression based on damage to noncomputational fire modeling tools consistent with the assessed as meeting the first target and/or manual guidance in Section 11.5.1.7.1 of NUREG/CR-6850. CAT I1. There is no suppression only for time to While there may be conservatisms associated with the impact to the Sb FSS-D3 damage of first target of 15 selection and application of particular fire modeling tools, application.
| |
| (CAT I) minutes. the fire modeling tools described in NUREG/CR-6850 are considered sufficiently accurate, despite any associated No specific fire modeling, conservatisms, for the fire modeling of all physical 4-13 calculations, or analysis were analysis units and scenarios to be sufficiently realistic done of the significant fire units rather than bounding.
| |
| analyzed in the quantification tasks. More analysis was The selection and application of fire modeling tools was included for the MCR with respect part of a larger iterative process for evaluating high risk to abandonment, however, there contributors. This process involved a team of still significant conservatism knowledgeable individuals with diverse expertise, remaining in the calculations such including fire modeling, circuit analysis, PRA, and plant as the below noted in BPN-PSA- operations. As described in Attachment 39 of BNP-PSA-080 'The sensitivity analysis 080, fire scenarios were evaluated based on total ODE presented in Appendix B impact and importance of individual fire cutsets or
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR,/F&O # Finding Resolution 5b Impact indicates that the fire growth rate groupings of fire cutsets. More detailed fire modeling was and the burning regime can one of several approaches (e.g., crediting conditional influence the predicted MCR circuit failure probabilities or operator recovery actions) abandonment times given a peak available to remove excessive conservatism and thereby heat release rate. However, to to achieve more realistic risk results. The determination to fully address these parameters in select and apply more detailed fire modeling tools was greater detail would require an based on expert opinion of the expected resulting analysis of individual cabinet improvement. If the chosen approach resulted in a less-enclosures and an assessment of conservative, more-realistic risk, then the risk the fire development and contributions of that scenario diminished and the next ventilation conditions for each most important cutset could be subjected to the process.
| |
| cabinet considered.' Therefore The process also provided for possible plant modification only capability Category I is to reduce the risk for the scenario to acceptable levels considered met. when conservatism was not either readily apparent or easily removed. Although repeated iterations tend to produce diminishing results, this process stopped upon meeting the goal of a FPRA model that provides realistic risk estimates with some reasonable margin to the requirements of RG 1.174.
| |
| BNP-PSA-086 Section 10 A quantitative evaluation of parametric uncertainty for Resolution of this FSS-E3 contains the identified sources of both ODE and LERF was performed as documented in finding provided the (CAT I) uncertainty in the fire modeling EVAL EC 296040, including a State of Knowledge required quantitative FSS-H5 scope. This evaluation was Correlation covering fire ignition frequencies, non- evaluation of (CAT I) limited to a qualitative evaluation suppression probabilities, conditional failure probabilities, uncertainty. This FS-9 of the identified uncertainties. No and fire bins. finding is sufficiently FS-9 statistical representations of the resolved for SR FSS-(CAT I/llI/Ill) uncertainty intervals was present, E3 and SR FSS-H5 to UNC-A2 therefore only Capability be assessed as (CAT 1/11/111) Category I was considered met. meeting CAT II.
| |
| There is no impact to The heat release rate, the the 5b application.
| |
| 4-14 shortest distance from the ignition source to the target and the fire
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact diameter are typically considered
| |
| - for statistical representation of uncertainty intervals. The remaining inputs of compartment geometry and ventilation characteristics are obtained from plant drawings are typically not subject to statistical uncertainty analysis.
| |
| BPN-PSA-080 calculation Section Following the methodology in NUREG/CR-6850, BNP- Resolution of this 6 evaluates the impacts of the PSA-080 calculation has been revised and the Multi- finding revised the MCA evaluations. In Section 6 Compartment Analysis does not assume a CCDP of 1.0 analysis for MCA.
| |
| only two MCA scenarios were not for any compartment in the MCA analysis. Compartment This finding is FSS-G6 screened, and required CCDPs were calculated based on actual localized target sufficiently resolved (A ) evaluation. For these two zones, sets for exposing compartments. for SR FSS-G6 to be (A ) the CCDPs were assumed to be assessed as meeting 1 and CLERP was assumed to be CAT IlI/ll. There is no 4-6 .1; therefore no specific impact to the 5b quantitative evaluations were application.
| |
| performed for these MCAs. As a result Capability Category IlIll1 is not met.
| |
| Existing active components A systematic review of the Level 2 progression for fire Resolution of this LE-E1 identified in the internal events impacts has been performed to the extent required by the finding performed a (NOT MET) models were considered in design of the core damage sequence models and Level 2 systematic review of F-1 component selection and cable model. This review has been documented as the Level 2 FQ1 routing. Quantification was Attachment 14 in the component selection calculation, progression for fire (CAT 1I/Ill11) performed using the existing BNP-PSA-085, Revision 2. impacts. This finding accident progression with no is sufficiently resolved noted changes as related to the for SR LE-E1 to be 4-17 affect of fire scenarios. Existing assessed as meeting modeled operator responses CAT 1I/1I/l1. There is
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact were evaluated for changes due no impact to the 5b to fire affects. ,The MSO application.
| |
| evaluation considered affects of LERF with respect to failure of containment isolation. However, no systematic review of the accident progression to determine if fire affects would impact the existing internal events accident progression was found.
| |
| Parametric uncertainties that are A quantitative evaluation of parametric uncertainty for Resolution of this QU-E3 associated HLR-DA, HR and IE both ODE and LERF was performed as documented in finding provided the (CAT I) are documented in BNP-01 87. EVAL EC 296040, including a State of Knowledge required quantitative QU-A3 However, the state of knowledge Correlation covering fire ignition frequencies, non- evaluation of (CAT II) correlation was not considered in suppression probabilities, conditional failure probabilities, uncertainty. This UN-l the evaluation of these and fire bins. finding is sufficiently UN-1 uncertainty evaluations, resolved for SR QU-(NOT MET) Correlation should be considered E3 to be assessed as FQ-A4 for fire events such as the fire meeting CAT II and (CAT I/llI/Ill) frequency, applied severity/HRR SR FQ-A4 to be split fractions, non-suppression, assessed as meeting circuit failure probabilities, etc. CAT I/Il/Ill. There is 4-18 no impact to the 5b application.
| |
| P-P-Al Section 3.2.1 and Attachment 1 of BNP-PSA-083 was revised to add Drywell/Torus, Spent Resolution of this (NOT MET) calculation BNP-PSA-083 have Fuel Pool, and VP1/VP2 to Global Plant Analysis finding revised areas been reviewed to examine the Boundary. The Spent Fuel Pool and Service Water Valve included in the global (CAT I/Il/Ill) process by which the Global Plant Pits were then qualitatively screened, while the plant analysis Analysis Boundary (GPAB) has Drywell/Torus were simply not quantitatively analyzed boundary and the PP-~C2 been defined in the BSEP FPRA. based on no fire being postulated in an inerted criteria for exclusion.
| |
| (NOT MET) Section 3.2.1 indicates that all atmosphere. This finding is areas that contained any sufficiently resolved equipment or cable credited in the The characterization of equipment as "risk significant" for both SR PP-Al
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding JResolution 5b Impact 5-1 FPRA were included, as well as was removed from the description of the criteria for and SR PP-C2 to be any area would require a plant excluding areas from the GPAB. The distances assessed as meeting shutdown. In addition, any area separating certain buildings of potential interests were CAT I/Il/Il1. There is that is adjacent to an area that added. no impact to the 5b would affect EPRA application.
| |
| cables/equipment or require a Distance from ABH to DGB is 32'.
| |
| shutdown is said to be included in the GPAB. All of these criteria are Distance from CTPH1 to DGB is 28'.
| |
| in agreement with PP-Al.
| |
| Distance from STORES to RB2 is 30'.
| |
| However, in Attachment 1, a number of buildings/areas are excluded from the GPAB because they do not affect "risk significant" equipment and they may not require a plant shutdown prior to the assumed threshold of 8 hours (described in Assumption 3.1.1.5). This process is consistent with the guidance provided for the Qualitative Screening Task (task 4) in Section 3.3, but is considered inappropriate for use at the PP stage of the analysis All other areas listed in the table in Attachment 1 should either be confirmed to contain no equipment or cables that are either:
| |
| : 1) credited in the FPRA (i.e., not just risk significant), or
| |
| : 2) capable of adversely impacting
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution 5b Impact plant response Additionally, the exclusion basis needs to include additional discussion for the following:
| |
| *Aux. Boiler House (ABH) - State that the closest building of concern is the DG building which is approx ___ ft away and will not be affected by an exposure fire in ABH.
| |
| * CTPH1 - Due to proximity to DG building, discuss-fire exposure potential.
| |
| * Fire house - Any fire alarm panels being affected?
| |
| * STORES - Address exposure to south side of the U2 rector building.
| |
| *VP1, VP2 - Not shown on the BGA boundary drawing.
| |
| ES-A5 There are a few fire-induced No change will be made to incorporate Finding 5-4 Resolution of this (CAT 1,1II) spurious events that were because further consideration of the listed spurious finding justified the ES-A6 screened, but could in fact either events revealed no additional fire impacts beyond what current approach.
| |
| (CAT I, II) cause a plant trip (or manual was already identified by the MSO Expert Panel. In There is no impact to ES-B2 shutdown) and impact equipment particular, contrary to Finding 5-4, the events described in the 5b application.
| |
| (CTII) that is credited for accident Items 1, 2, 4, 5, and 6 neither cause an automatic plant (CT11) mitigation in the FPRA: trip nor require a Tech Spec mandated manual shutdown ES-D1 in less than the 8-hours assumed for treatment as a f ire-(CAT I/Il/Ill) 1) spurious start / injection by induced plant trip.
| |
| PRM-B9 RCIC. This was screened from (CAT I/l1/l1l) the FPRA and an initiating event Although the event described in Item 3 could either cause because it was assumed that no an automatic plant trip or prompt the Operator to initiate a
| |
| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions SR/F&O # [Finding -
| |
| JResolution J 5b Impact plant trip would occur. However, a manual scram (depending on the number of individual 5-4 fire-induced RCIC start would control rods that initially scrammed), the RPS scram likely only be caused if the fire signal itself would shortly close the SDV vent and drain damage was significant enough valves, at which point the scenario would most resemble to cause ROIC inoperability. a previous addressed turbine trip.
| |
| Assuming no plant shutdown may be non-conservative. With regard to the suggested possible resolutions:
| |
| For items 1-3, since Section 3.3.2.1 of the MSO report
| |
| : 2) spurious start/Iinjection by (Attachment 4 of Calculation BNP-PSA-085) already HPCI. This was screened from documents significant operator experience for members the FPRA and an initiating event of the MSO Expert Panel, there is little marginal benefit in because it was assumed that no citing additional operator interviews for support.
| |
| plant trip would occur. However, a For item 4, the only clarification necessary would be to fire-induced HPCI start would note that the item is incorrectly premised on core spray likely only be caused if the fire being able to inject at high RCS pressure.
| |
| damage was significant enough For item 5, the identification in the MSO report of a to cause HPCI inoperability. restricting orifice with a 0.105 inch bore should already be Assuming no plant shutdown may sufficient documentation that the HPCI drain pot line to be non-conservative. the condenser does not constitute a steam flow diversion.
| |
| : 3) MSO item C11 -2e. This MSO For item 6, since an automatic plant trip or manual drains the RPV through the SDV shutdown is required to drop RPV to below that needed vent and drain. The exclusion of for condensate injection, a plot of RPV pressure over time this event from the FPRA is is not needed to invalid this MSO (i.e., spurious based on the fact that the condensate injection with RPV pressure below 500 psig) suppression pool inventory as an initiating event and would add nothing to the depletion is slow and would not evaluation, in the MSO Report (Attachment 4 of reach a low enough level in 24 Calculation BNP-PSA-085), of equipment credited for hours to require a plant trip. post-trip accident mitigation.
| |
| However, it may be nonconservative to assume that there is no chance of a plant trip due to this uncontrolled loss of RCS inventory.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # jFinding ]Resolution j 5b Impact
| |
| : 4) MSO item E21-O01. This MSO describes spurious actuation of the core spray pumps and spurious operation of the injection valves. This event can cause flooding of the main steam lines, which can subsequently cause failure of the turbine-driven RCIC/HPCI pumps (and FW, which is not modeled). The exclusion justification says that high-pressure injection is 'not credited after depressurization',
| |
| so there is no way to model the event. However, if spurious CS pump operation occurred at high RCS pressure and the main steam lines were flooded, HPCI and RCIC should be impacted because there is still potential for crediting their high-pressure injection.
| |
| : 5) MSO item E41 -2w. This MSO describes the unisolated drain of HPCI to the main condenser via spuriously opened AOVs. Two of the three AOVs in series have been locked open, so this scenario only requires one AOV to open (on loss of instrument air or hot short). This event is excluded based on an installed
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact flow-limiting orifice, but there is no technical discussion of the flow limitation to adequately justify why the flowpath is not a valid diversion.
| |
| : 6) MSO item N21-2ai. This MSO describes RPV overfill due to condensate injection once RPV pressure is <500psi. The exclusion justification states that it is unlikely that RCIC/H PCI operation alone would not depressurize the RPV to 500psi in one hour. However, ROIC and HPCI are credited for injection for much longer than 1 hour, at some point RPV pressure may be reduced to allow condensate injection, which could potentially fail HPCI/RCIC.
| |
| Some of the exclusion bases for Section 3.4.3 of BNP-PSA-083 was revised to include Resolution of this the BSEP historical fire events additional discussion of the plant history and corrective finding reviewed the should be strengthened to actions concerning fires related to the heater drain pumps fire events in question IGN-A4 support the conclusion that the (Items #1, #2, #5, and #7). The appropriate exclusion of and revised the (CAT 1,II) use of generic ignition frequency Item #3 as being outside the GPAB was confirmed. The calculation, as IGN-B4 data is appropriate: appropriate exclusion of Item #4 and Item #6 as being not necessary, to (CAT VIIl/Ill) potential challenging was confirmed. Some further document the basis
| |
| : 1) FR 88-006: A heater drain clarification of the documentation for Items #3, #4, and #6 for considering them pump ignited and required 3 CO2 was considered but judged unnecessary at this time. as being not 5-8 extinguishers at power. potentially Approximately 2 quarts of oil were challenging. There is burned. This appears to be no impact to the 5b potentially challenging, application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # fFinding jResolution 5b Impact
| |
| : 2) FR 90-002: A heater drain pump ignited at power and required 'several' extinguishers.
| |
| Fire was fueled by pump oil, caused a fire alarm, and resulted in -$64k worth of damage. This appears to be potentially challenging.
| |
| : 3) FR 94-007: A OWOD pump ignited at power and required offsite fire department response.
| |
| IF this was not dismissed in PP, this could be a potentially challenging fire.
| |
| : 4) ACR 94-01 488: A fire in a Rad Waste control room panel at power required a fire extinguisher. Fire caused loss of SFPC, which appears to be potentially challenging.
| |
| : 5) ACR 97-1136: A heater drain pump ignited at power and was secured to extinguish the fire in response to the fire alarm. This appears to be potentially challenging.
| |
| : 6) ACR 98-651 : A cable fire started in a manhole at power due to water intrusion and
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # JFinding Resolution 5b Impact corrosion. Fire was self-extinguishing, but cable damage was reported.
| |
| : 7) NCR 24699: A heater drain pump ignited an oil fire, which caused a fire alarm and required 002 extinguishers while at power. A condensate system transient resulted and an unusual event was declared due to a duration of >10 minutes. This appears potentially challenging.
| |
| There is no record of a review A systematic review for modeling inconsistencies Resolution of this being performed to confirm that finding performed a associated with fire impacts was performed and has the FPRA modeling is consistent systematic review for been documented in various sections of BNP-PSA-085, from event sequence to system modeling with conclusions consisting of statements of acceptability model or that with operational inconsistencies QU-D2 or descriptions of the required changes.
| |
| (NOT MET) characteristics. Since the FPRA associated with fire Section 3.3.1.4 of BNP-PSA-085 describes the review of model is largely based on the impacts. This finding QU-F3 the PRA Internal Events model Accident Sequence (CAT I) internal events model, this is is sufficiently resolved Notebook (i.e. BNP-PSA-029) and Level 2 Accident assumed to be a relatively for SR QU-D2, SR F'Q-E1 Sequence Notebook (i.e. BNP-PSA-049). These PRA insignificant source of potential EQ-El, and SR FQ-(NOT MET) notebooks address the plant response and the event model inaccuracy. However, a F1 each to be EQ-Fl trees developed for that response. The review determined review does need to be assessed as meeting that the EPRA will be maintained as part of the Internal (NOT MET) performed to confirm that fire- CAT I/ll/l1l and for SR Events PRA Model of Record and concluded that the use specific modeling considerations QU-F3 to be of the same PRA models as for the internal events have not created any assessed as meeting 5-13 sequence quantification ensures that inconsistencies between CAT IlI/ll. There is no interdependencies are modeled consistently and sequence and system modeling, impact to the 5b appropriately. Because the BSEP PRA uses functional or between the EPRA model and tops for the event trees, and functions are modeled by application.
| |
| actual plant operational practices. initiating events and by system models, a detailed sequence by sequence review is not required and would
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding [Resolution 5b Impact provide no benefit. The fire events cannot change the modeled functions, and the conclusion is always that the accident sequence logic is adequate because the functions are always the same: reactivity control, RPV integrity, inventory control, pressure control, decay heat removal, and containment integrity.
| |
| Section 3.3.1.4 of BNP-PSA-085 also describes the review of applicable initiating events (i.e., BNP-PSA-032) and the relevance to the fire model (i.e., BNP-PSA-085, Attachment 8). Consideration of possible additional initiating events that might be unique to the FPRA is documented in Table 3-2 of BNP-PSA-085.
| |
| Attachments 3 and 4 of BNP-PSA-085 document the review and disposition of various fire-induced MSOs postulated by plant and industry personnel to have potential impact on mitigation functions and systems. This review resulted in certain FPRA model changes as documented in Attachments 9 and 12 of BNP-PSA-085 and included the creation of a simplified bypass event tree for a main steam isolation valve (MSIV) MSO, as described BNP-PSA-085, Section 3.3.1.4 and Attachment 13.
| |
| For the Level 2 review, a detailed review of the containment isolation is performed in BNP-PSA-085, Attachment 6. Subsequently the review of the fire impact on the Level 2 accident sequences and phenomenological events was performed as documented in Attachment 13 of BNP-PSA-085.
| |
| QU-F3 A review of the cutset review Non-significant cutsets were reviewed and the results are Resolution of this (A ) documentation indicates that the documented in Attachment 39 of BNP-PSA-080 (i.e., finding reviewed non-(A ) vast majority (if not all) of the Change Package BNP-0235). significant cutsets.
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| QU-D5 reviewed cutsets are from This finding is (NOT MET) significant scenarios, almost ______________________sufficiently resolved
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact EQ-El exclusively with CCDPs of 1.0. for SR QU-D5, SR (NOT MET) Many of these CCDP cutsets EQ-El, and SR FQ-E-l have only a single cutset (other Fl each to be (NOT MET) applicable cutsets are truncated). assessed as meeting At the current stage of the BSEP CAT I/Il/Ill and for SR EPRA development, this is not an QU-F3 to be 5-4 unreasonable characteristic of the assessed as meeting cutset reviews. CAT IlI/ll. There is no impact to the 5b However, a lack of review of non- application.
| |
| significant cutsets precludes
| |
| ___________meeting this SR.
| |
| Although a limited identification of The correctness and reasonableness of the FPRA Resolution of this fire CDF contributors has been modeling were reviewed based on detailed cutset reviews finding documented performed, the types of for individual scenarios, such as that documented the review of FPRA QU-E2 contributors is limited and there is in Attachment 39 of BNP-PSA-080. The results of the results and the (CAT 1/1l/I1l) little or no discussion of the risk Fire PRA have been reviewed and top contributors by identification of risk QU-F3 insights gained from the ignition source, including transient and fixed, and insights. This finding (CAT I) contributor identification. compartment have been identified and listed in Section is sufficiently resolved Q-63.4 of Attachment 38 to BNP-PSA-080, by cutset for SR QU-D7, SR QU-D6I For example, the following probability, and the top contributors to both ODE and EQ-El, and SR FQ-(CATuin I)db nihtu, LR.Te ikcnrbtosadrs mprac vns 1ec ob QU-D7 contribtion oulbee idnsgtiful, LERd Theairsk conributosesd anddsrisk imoanc d eventso El3 assedach tomeetn (NOT MET) anoaentbe ietfe: adfAtailurest wer asBN-Seed , asnecrbd in Sctuerniong33 ssessed/I asdmeeting FQ-Eof-Attachmentn38atoidentPsAq8encsand includedorankingseCATiI/Il/Illand forQSR (NOT MET) - risk significant operator actions -Fire Compartments F3 to each be EQ-Fl performed inside the main control -Fire Scenarios assessed as meeting (NOT MET) room -Fire Accident Sequences CAT IlIll1. There is no
| |
| - risk significant operator actions -Containment Failure Types (i.e., LERE only) impact to the 5b performed outside the main -Operator Actions application.
| |
| 5-15 control room -Fire Induced Equipment Eailure Modes
| |
| - contribution to fire CDF from -admCmoetFiue transient ignition sources
| |
| - contribution to fire ODE from -Systems
| |
| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions SR/F&O # {Finding -
| |
| Resolution 5b Impact fixed ignition sources -Component Type Failures
| |
| - significant spurious actuation For each of these categories other than Containment events Failure Types, the top contributors were ranked for both
| |
| - significant random failure events CDF and LERF according to the percent contribution to (i.e. non-fire), including common risk. For Containment Failure Types, the assessment only cause failures considered the contributions to LERF. The importance
| |
| - the REDUCTION in ignition ranking results in each of these ranking categories are frequency contribution to fire CDF generally used in addressing which portions of the FPRA due to the extensive use of the model need further refinement.
| |
| conditional plant trip probabilities Insights from importance analysis were used to review the correctness and reasonableness of the FPRA modeling Additionally, the importance of by comparing the results against what is normally components and basic events understood about plant response. Attachment 38 of BNP-were not reviewed to determine PSA-080 provided the following insights for reviewing the that they make logical sense (QU- correctness and reasonableness of the FPRA model:
| |
| D7). -The MCR and cable spreading rooms dominate risk contributions from Fire Compartments;
| |
| -Scenarios involving fixed ignition sources, rather than transient combustibles, are major contributors;
| |
| -All transient and station blackout (SBO) sequences for fire result in either loss of makeup events or loss of decay heat removal events that result in a loss of makeup;
| |
| -With emergency power blackout associated with the dominant cause of core damage, failures of fail-safe containment isolation valves may not contribute as much to LERF as presented;
| |
| -Control room abandonment for habitability is one of the more important operator actions;
| |
| -Lack of a specific method for evaluating fire-induced instrument faults is evident in the results;
| |
| -Random component failure rankings show test and maintenance unavailability is important to fire risk;
| |
| -The plant system that contributes most to fire risk is the
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b impact AC power system, with AC breakers as contributing components.
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| With respect to identifying the The correctness and reasonableness of the FPRA Resolution of this contributors to fire LERF, the modeling were reviewed based on detailed cutset reviews finding documented following contributors are for individual scenarios, such as that documented the review of EPRA considered: in Attachment 39 of BNP-PSA-080. The results of the results and the Fire PRA have been reviewed and top contributors by identification of risk
| |
| - contributions from fire scenarios, ignition source, including transient and fixed, and insights. This finding LE-Fl MCA abandonment, and the compartment have been identified and listed in Section is sufficiently resolved (NOT MET) multi-compartment analysis. 3.4 of Attachment 38 to BNP-PSA-080, by cutset for SR LE-F2, SR FQ-
| |
| - compartments with >1% fire probability, and the top contributors to both CDF and El, SR EQ-Fl and SR LE-F2 LEFLERF. The risk contributors and risk importance events UNC-A1 each to be (NT ET -ignition sources with >1% LERE and failures were assessed, as described in Section 3.3 assessed as meeting I-E-G3 of Attachment 38 to BNP-PSA-80, and included rankings CAT I/Il/Ill and for SR (CAT I) No identification of plant damage in the following categories: LE-F1 and SR LE-G3 UNC-A1 states or containment failure -Fire Compartments each to be assessed (NOT MET) modes was identified, which is -Fire Scenarios as meeting CAT IlI/ll.
| |
| EQ-El required for CCI. To meet CClI, -Fire Accident Sequences There is no impact to (NOT MET) additional identification of -Containment Failure Types (i.e., LERE only) the 5b application.
| |
| EQ-El significant fire LERE contributors -Operator Actions (NOT MET) is required, as discussed in the -Fire Induced Equipment Failure Modes SR. -Random Component Failures Within the scope of fire LERF -Systems 5-16 contributors that have been -Component Type Failures identified, it is not apparent that a For each of these categories other than Containment review for 'reasonableness' has Failure Types, the top contributors were ranked for both been performed. CDF and LERF according to the percent contribution to risk. For Containment Failure Types, the assessment only For example, 98.1% of Unit 2 fire considered the contributions to LERF. The importance LERE is due to fires in the Unit 2 ranking results in each of these ranking categories are
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding ]Resolution 5b Impact main control room. Although this generally used in addressing which portions of the FPRA is identified in table 11-2 of BNP- model need further refinement.
| |
| PSA-080, there is no discussion Insights from importance analysis were used to review the of this considerable contribution correctness and reasonableness of the EPRA modeling including whether or not it is by comparing the results against what is normally considered reasonable. Notably, understood about plant response. Attachment 38 of BNP-the Unit 1 MCR contributes -60% PSA-080 provided the following insights for reviewing the of Unit 1 fire LERF, and no correctness and reasonableness of the FPRA model:
| |
| discussion of this asymmetry is -The MCR and cable spreading rooms dominate risk provided contributions from Fire Compartments;
| |
| -Scenarios involving fixed ignition sources, rather than transient combustibles, are major contributors;
| |
| -All transient and station blackout (S60) sequences for fire result in either loss of makeup events or loss of decay heat removal events that result in a loss of makeup;
| |
| -With emergency power blackout associated with the dominant cause of core damage, failures of fail-safe containment isolation valves may not contribute as much to LERF as presented;
| |
| -Control room abandonment for habitability is one of the more important operator actions;
| |
| -Lack of a specific method for evaluating fire-induced instrument faults is evident in the results;
| |
| -Random component failure rankings show test and maintenance unavailability is important to fire risk;
| |
| -The plant system that contributes most to fire risk is the AC power system, with AC breakers as contributing components.
| |
| LE-G2 Assumptions for the quantification A quantitative evaluation of parametric uncertainty for Resolution of this (NOT MET) task are documented in Section both CDF and LERF was performed as documented in finding documented LE-F3 3.3 of BNP-PSA-080. General EVAL EC 296040, including a State of Knowledge the review of sources (NOT MET) sources of uncertainty are Correlation covering fire ignition frequencies, non- of uncertainty for LE-G4 discussed in Section 8.4. These suppression probabilities, conditional failure probabilities, LERF. This finding is
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # [Finding Resolution j 5b Impact (NOT MET) sources include: and fire bins. sufficiently resolved UNC-A1 for SR LE-F3, SR LE-(NOT MET) - ignition frequencies As evaluated in BNP-PSA-080, Attachment 38, the most G2, SR LE-G4. SR UNC-A2
| |
| - HRRs significant area of epistemic uncertainty with regard to UNC-A1, SR EQ-El, (CAT I/Il/Ill)
| |
| - target selection LERE is in circuit analysis, specifically as related to and SR EQ-El each
| |
| - damage time spurious operation of containment isolation valves. to be assessed as EQ-El - time to HGL Lacking specific guidance in NUREG/CR-6850 for the meeting CAT I/Il/Ill.
| |
| (NOT MET) - fire effects treatment of low voltage instrumentation loops and the There is no impact to EQ-Fl - suppression grounding or clearing hot shorts in DC circuits, these the 5b application.
| |
| (NOT MET) - circuit analysis events were assigned a value of 1.0 in the BSEP FPRA.
| |
| -HRA The resulting cutsets are dominated by signal failures
| |
| - quantification (including tools) causing valve spurious operations or primary containment 5-18 isolation valves (PCI Vs) remaining spuriously open, even These sources of uncertainty are though their design is to fail safe in the closed/isolated valid in the fire LERF and fire position. A more realistic assessment of these affects ODE quantifications, but there are would greatly reduce LERF.
| |
| no additional sources of uncertainty that are applicable to The sources of aleatory uncertainty were evaluated for the fire LERE calculation. Change LERF, and a detailed results analysis was performed for package BNP-O0187 provides fire LERF as documented as Attachment 38 to the BNP-PSA-CDF importance measures and a 080 and supplemented in EVAL EC 292418. This analysis statistical analysis of fire CDE includes evaluation of parametric uncertainty for a uncertainty, but does not address combined LERF solution and various importance fire LERF. evaluations for the same solution, including the State of Knowledge Correlation. In particular, parametric uncertainty was evaluated for the following:
| |
| (a) Fire scenario event frequencies (i.e., initiators),
| |
| (b) Component failure probabilities (i.e., random faults and hot short probabilities),
| |
| (c) Component maintenance unavailability, (d) Human error probabilities, (e) Common cause failures, and (f) Recovery Actions (i.e., main control room abandonment from environmental causes)
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The types of fire data reviewed for SOKO included: fire ignition frequency, non-detection probabilities, non-suppression probabilities, heat release rate severity factor/split fraction, and circuit failure probabilities.
| |
| The top contributors to LERE were ranked for: fire compartments, fire scenarios, accident sequence types, containment failure types, operator actions, fire-induced equipment failure modes, component types, and failure
| |
| _________________________groupings.___________
| |
| There is no definition established A discussion of "significance" in terms of the definitions Resolution of this for 'significance' related to basic described in ASME/ANS-Ra-Sa-2009 Section 1-2 has finding added a LE-G6 events, cutsets, accident been added to Section 8 of the Quantification Calculation discussion of (NOT MET) sequences, or any other facets of (i.e., BNP-PSA-080). significance to QU-F6 the fire PRA results. documentation. This (NOT MET) finding is sufficiently FQ-F1 resolved for SR LE-(NOT MET) G6, SR QU-F6, and SR FQ-F1 each to be assessed as meeting 5-19 CAT I/Il/Ill. There is no impact to the 5b application.
| |
| Thedefciet sb-equremntsofAttachment 13 was added to BNP-PSA-085, Revision 2, Rslto fti this SR are detailed below, to address items A, B, D, E, F, and part (i.e., equipment, finding added LE-G2 A) No documentation was containment failure modes and phenomena) of C. documentation to (NOT MET) provided of plant damage states / address the FQ1 attributes, although this can be The remainder of item C (i.e., fire-specific human actions requirements. This FQ-F1 considered covered by general considered in the fire LERF sequence development) was finding is sufficiently (NOT MET) references to the internal events addressed in Section 4.2.3 and Table 5.1 of BNP-PSA- resolved for both SR PRA model. 084, Revision 2. LE-G2 and SR EQ-Fl 5-20 B) There is no documentation of F&O 5-18 sufficiently addresses item H (i.e., LERF- to be assessed as how accident sequences were related uncertainty). Resolution was completed as part of meeting CAT I/Il/lll.
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| binned into plant damage states, F&O 5-18. There is no impact to but since the fire LERF model is the 5b application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # ]Finding ]Resolution 5b Impact based on the internal events LERF model, references to the internal events PRA can account for this.
| |
| C) There SHOULD be discussion of the fire-specific human actions and equipment considered in the fire LERF sequence development. Containment failure modes and phenomena could be referenced to the internal events documentation D) There is no discussion of fire-specific factors influencing containment challenges and containment capability.
| |
| E) Containment capacity analysis could be covered by a reference to the internal events LERF model. No fire-specific impacts are expected.
| |
| F) A discussion of fire-specific impacts on the accident sequences identified in the containment event trees should be provided.
| |
| H) The model integration process is described in Section 4.9 of BNP-PSA-080. There are no fire LERF-related uncertainty (F&Q 5-
| |
| : 18) or sensitivity analyses provided.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution J 5b Impact There is evidence that the Evidence of the collective review of electrical coordination Resolution of this existing electrical coordination with supporting analysis for breakers, power supplies and finding reviewed and analysis was reviewed and cables was documented in BNP-PSA-080 as: documented electrical refined (e.g. BNP-O0157). Specific Attachment 13 (i.e., Change Package BNP-0157) coordination with documentation should be Attachment 36 (i.e., Change Package BNP-0218) supporting analysis provided of this review. There is Attachment 37 (i.e., Change Package BNP-0224) for breakers, power no evidence that power supplies Attachment 41 (i.e., Change Package BNP-021 5) supplies and cables.
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| credited in the fire PRA were Attachment 42 (i.e., Change Package BNP-0217) This finding is reviewed to confirm that they Attachment 43 (i.e., Change Package BNP-0223) sufficiently resolved were addressed by existing for SR CS-B 1 to be overcurrent calculations. assessed at meeting Attachment 36 of BNP-PSA-080 identifies two raceways CAT IlIll1 and SR CS-that could not be routed on drawings. However, C4 to be assessed as Section 5.d of Attachment 36 describes how a reasonable meeting CAT I/Il/Ill.
| |
| CS-B1 approximation for the fire zones that the raceways There is no impact to (CAT I) traverse was possible based on the raceways adjacent to the 5b application.
| |
| CS-C4 the raceway in question, the cable start and end (NOT MET) equipment, and general plant layout knowledge.
| |
| Attachment 37 of BNP-PSA-080 identifies three other 6-1 raceways that could not be routed. The raceways that are missing information are located in the Unit 2 electrical equipment room (i~e., Control Building 49'). The raceway for 2-2A-120V involves a cable running between two panels in adjacent rows in the Unit 2 electrical equipment room (i.e., Control Building 49'). The two raceways for 2-2D-120V involve a cable running from a panel in the Unit 2 electrical equipment room (i.e., Control Building 49') to a panel in the Unit 2 cable spreading room (i.e., Control Building 23'), which is where the two other raceways are known to be located. These three raceways are identified as a source of uncertainty in Section 3.6.6 of BNP-PSA-080, and the risk associated with their assumed failure is qlualitatively addressed as a
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution 5b Impact non-conservative assumption (i.e., Section 3.1.3.44) that is likely mitigated in the HGL scenario by other failures for the respective power supplies.
| |
| All panels modeled in the FPRA were included within the scope of the breaker coordination study, as described in Attachment 42 of BNP-PSA-080.
| |
| Passive fire barriers with a fire Section 11.5.4 of NUREG/CR-6850 requires postulating Resolution of this resistance rating are credited in the failure of only one fire barrier, and selecting the worst finding justified the the multicompartment analysis. case value for those applicable to the Fire Compartment current approach.
| |
| FSS-G4 The failure rates used are those is conservative. There is no guidance for summing the This finding is (CAT I) prescribed in NUREG 6850, probabilities for individual elements over an entire wall to sufficiently resolved however, the worst case value for get probability of wall failure. Walkdowns were performed for SR FSS-G4 to be failure probability of the barrier is to gather the targets and barriers between the exposing assessed as meeting 6-4 used. and exposed compartments. The worst case barrier CAT II. There is no failure probability was applied to all local targets between impact to the 5b two adjacent compartments. The results of this analysis application.
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| is included in Attachment 7 of BNP-PSA-080.
| |
| Screening methodology is As described in Attachment 7 to BNP-PSA-080, plant Resolution of this provided in BNP-PSA-080, walkdowns were performed to identify targets in the revised the calculation Section 6.0. exposed compartments near the barriers separating the to include the exposing and exposed compartments. The localized localized damage in FSSG2 However: the MCA screening did damage in the adjacent compartment near barriers for all the adjacent (CAT I/Il/Ill) not consider the impact of compartments that screened out and for compartments compartment near possible localized effect (e.g., where MCA was performed but did not achieve a HGL in barriers. There is no damage to equipment) near the combined compartments was included. The localized impact to the 5b 6-5 penetrations and barriers, targets of the adjacent compartment were added to the application.
| |
| HGL evaluation for the exposed compartment.
| |
| In addition, a screening value was used without justification and the cumulative risk for the screened scenarios was not evaluated.
| |
| | |
| Tab&O#lenin 3.RE iePe eiwFnig bervautions Resolution SRIF&O # Finding Resolution 5b Impact Conditional failure probabilities No non-conservative application of conditional failure Resolution of this were assigned to selected cables probabilities has been identified for an off-scheme cable. finding justified the per the methodology identified in current approach.
| |
| BNP-01 37, which is based on the For safe shutdown, any failure of an associated circuit This finding is Chapter 10 tables in NUREG also fails the main component. This is conservative, in sufficiently resolved 6850. However the BSEP that (typically) only one or two of an associated circuit's for SR CF-Al to be methodology for determining the cables actually affect the primary component. When assessed as CAT component level spurious applied to the Fire PRA, this method of including IlIll1. There is no operation probability, as identified associated circuits created far too many false failures, impact to the 5b in BNP-PSA-080 Section 4.3.4 and therefore associated circuits are not always linked to application.
| |
| CF-Al and 4.6.1.2.4, is to use the worst the primary component as shown in FSSPMD. In almost (CAT I) case spurious operation all cases, the associated circuits are modeled separately probability of all affected cables as primary components in the fire PRA fault tree. In this without regard as to whether the manner, cable damage to the associated circuit is 6-7 cables in question are primary captured within the fault tree, and will cause cascading scheme or off-scheme cables. failures based on the model. In addition, key interlocks Per FAQ 08-047, off scheme that can have an impact on the Fire PRA are included in cables and cables with alternate the model. Therefore it can be determined that off-source breakers must be scheme cables are included.
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| identified and, when combined with on-scheme cables, an Additionally, many times, although they are included, the exclusive OR must be used. failure probability may be 1.0, and appear to be Spurious events of high unanalyzed. In assigning the fault probabilities for importance that had spurious Brunswick, specific basic events were identified by PRA.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact operabilities applied were Fault probabilities were assigned to the on-scheme reviewed and found to have no cables that could affect the basic event of concern. The off-scheme cables, therefore Cat I values assigned represented the best-estimate as shown is considered met. in the tables in Chapter 10 of NUREG/CR-6850. These fault probabilities were in general, only applied to control circuits. A loss of power that results in the failure of a basic event could occur due to a short to ground, and since the fault probabilities provided in NUREG/CR-6850 only apply to hot shorts, a probability of 1.0 would be assigned. Similarly, instrumentation cables are assumed to fail with a probability of 1.0 since they have not been specifically tested. However, since they can fail either high or low, a split fraction may still be applied to the functional response to the cable fault. Since many of the associated circuits are tied to instrumentation, not performing a fault probability analysis on such circuits has no impact on the PRA results since the failure would be an assumed value of 1.0, and no advantage would be gained.
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The wind PRA model should External exposed components are failed through the use This issue has been include only those SSCs of a flag file. Components that reside in turbine building assessed and addressed by the fragilities are assumed protected from wind hazard to the same changes will be analysts, and trim the rest from extent as the loss of offsite power fragility. Most of applied to the model.
| |
| WP-3 PRA. components in turbine building are lost with the loss of There is no impact to (NOT MET) offsite power which is much more fragile than the turbine the 5b application.
| |
| building. Exposed components have been assessed and WPR-A3-0 1 dispositioned accordingly. Components fed by offsite power (circulating water pumps and startup transformers) are assumed as rugged as offsite power source that powers them. The component Basic Events failed by wind events have been identified.
| |
| From the current documentation, For long duration high wind events (hurricanes) operators This issue has been it is not apparent why only 3 will be stationed in the turbine building to perform certain assessed and HEPs were selected for multiplier HRA events (outlined in procedure 0AOP-13). For short changes will be effects from high winds, even duration events (high straight line and tornado winds) the applied to the model.
| |
| WPR-A5 though Attachment 13 shows event will end before the human action is needed. Since There is no impact to (NOT MET) many HEPs for which multipliers storms are not uncommon events, the performance the 5b application.
| |
| WPR-A8 would apply. shaping factors following the event should be minimally (NOT MET) impacted. The only actions that should be prohibited are events outside the control room, turbine building, reactor WPR-A5-O01 building or diesel building that occur within the first hour.
| |
| The model has been updated to fail short term external actions since they may not be viable and removes multipliers on the HEPs.
| |
| WPR-A5 There are two potential errors in Upon further evaluation it was determined that short term This issue has been (NOT MET) utilizing the multiplier criteria in external actions should be failed since they may not be assessed and WPR-A8 Table 4. First, ex-control room viable. changes will be (NOT MET) actions that do not traverse applied to the model.
| |
| through areas impacted by winds There is no impact to and performed in areas not the 5b application.
| |
| WPR-A5-02 impacted by winds do not need to
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact be considered at a higher failure probability than in-control room actions. This is potentially conservative.
| |
| Second, the criteria in Table 4 do not recognize the possibility that some ex-control room actions may have a guaranteed failure (e.g., an action that requires going outside in a severe hurricane). The maximum increase in the HEPs in Table 4 is a factor of 30. This is potentially non-conservative.
| |
| With high failure probabilities, The product of the high wind initiator and the given No further action there is the potential for success component's fragility are applied as direct failures to all of required. Therefore, branches of the event tree to be the applicable SSCs (product of failure frequency and there is no impact to overestimated. The quantification initiator frequency is very small) as such the assumption the 5b application.
| |
| WPR-A9 engine applies a min-cut upper of small probabilities maintains the validity of the min-cut (NOT MET) bound approximation of the point upper bound estimate to provide accurate results. As estimate, at higher wind intervals such, there is very little impact from not modeling the WPR-A9-01 the fragility values approach 1.0 complimentary success state.
| |
| and mmn-cut upper bound estimate is not sufficient to provide accurate results.
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact If system recoveries (e.g., service Upon further evaluation it was determined that recovery This issue has been water recoveries) are credited in actions that are not feasible due to equipment damage assessed and the WPR-A1 1 the model, then their potential to should be failed. The rule recovery file and flag file was changes will be (CAT I) be impacted by the high winds updated to reflect the changes made to this HRA applied to the model.
| |
| conditions needs to be evaluated, assessment. There is no impact to WPR-A1 1-01 the 5b application.
| |
| The requirement is to assess the Accident sequences have been reviewed and impacts of There is no impact to accident sequences. While the assumptions have been considered. This is a the 5b application.
| |
| ability to assess the sequences documentation issue.
| |
| WPR-B1 has been developed, there is no (CAT I) documentation of the high winds (CAT I) quantification. It is also beneficial WPR-B -01 to review the results and consider the impact of any conservative assumptions that drive the results.
| |
| The requirement is to account for Error factors have been estimated from the range on the This issue has been the uncertainties in each of the hazard curves and an error factor of 10 has been assessed and inputs and for all important assumed for other basic events. The error factors have changes will be WPR-B2 dependencies and correlations. been added to the .RR file. This Finding pertains to applied to the model.
| |
| (CAT I) This has not been performed. documentation of this issue. There is no impact to the 5b application.
| |
| WPR-B2-01
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact The requirement is to document This Finding pertains to a documentation issue. Additional the specific adaptations to the documentation internal events PRA to produce delineating the specific WPR-C2 the high-wind PRA. This has not adaptations to the (NOT MET) been performed. internal events PRA model will not impact WPR-C2-01 the insights and results used to support the 5b application.
| |
| Sources of uncertainty and This Finding pertains to documentation of sources of Further documentation assumptions must be collected to uncertainties and assumptions in the model. of this issue does not meet the SR. It is also beneficial impact the insights (NOT-MET to characterize and assess the and results used to (OME) impact the sources of uncertainty support the 5b WPR-C3-01 may have on the model and application.
| |
| results.
| |
| XFPR-A3 The requirement is to ENSURE Evaluation of the potential external flooding impact in Resolution of this (NOT MET) the PRA models reflect external operator actions, how operator actions that are not finding evaluated and XFPR-A5 flood-caused failures. To provide credited were added to the PRA model, and a fragility documented potential (NOT MET) this assurance documentation is analysis for equipment was addressed. Two human external flooding XFPR-A8 needed for the systematic review reliability actions were not considered feasible: OPER- impact in operator (NOT MET) for potential impacts of external SWDISCHX and OPER-RESOSP. actions. There is no XFPR-A10 flooding. impact to the 5b (NOT MET) application.
| |
| XFPR-A3-01__________
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The likelihood of common-cause Potential for common cause failure due to large debris Resolution of this failures are an important related to external flooding was incorporated into the finding documented dependency. For example, the documentation and analysis. However, the fact that the the potential for XFPR-A7 clogging of intake structures and plant will not be at 100% power during severe weather common cause failure (NOT MET) other flow paths by debris related conditions, would minimize the amount of debris being due to large debris to the flooding should be sucked into the canal and subsequently to the intake related to external XFPR-A7-01 considered, and the walkdown is structure as the circulating water pumps would not be at flooding. There is no a means to ensure that this issue full power since the unit would be in shutdown mode. impact to the 5b has been evaluated properly. application.
| |
| If system recoveries (e.g., service System recoveries consideration and their potential to be Resolution of this water recoveries) are credited in impacted by the external flooding conditions were finding evaluated the model, then their potential to evaluated. The BSEP model does not consider any repair system recoveries XFPR-A1 1 be impacted by the external work after component failure. All failed systems are not potentially impacted (CAT I) flooding conditions needs to be considered recoverable during the transient, including by external flooding.
| |
| evaluated. loss of off-site power (LOOP). There is no impact to XFPR-A1 1-01 the 5b application.
| |
| The requirement is to document Modifications to the BSEP model include: A flag (FL- Resolution of this the specific adaptations to the EXTELOOD) to enable the external flooding model and finding documented internal events PRA to produce two new initiating events %EXTFLI (under gate the specific the external flood PRA, and to ACPTEEXTFLOOD-L) and %EXTFL_2 (under gate adaptations to the document the final results as well ACPTEEXTFLOOD-S), failing the LOOP recovery, internal events PRA to XFPR-C2 as selected intermediate results. %EXTFL_2 is a below design basis initiating event that produce the external (NOT MET) This has not been performed. models a 20 ft still water flood event and has an initiating flood PRA. There is event frequency of 7.4E-04/yr. %EXTFL_1 is a beyond no impact to the 5b XFPR-C2-01 design basis initiating event that models a 23 ft still water application.
| |
| flood event and has an initiating event frequency of 5E-05/yr. Both events fail the switchyard, the electric and diesel firewater pumps, and the circulating water pumps, except that %EXTFL_1 also fails the emergency diesel generators. However, as there is a very large degree of
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding ]Resolution 5b Impact uncertainty related to the frequency and duration of the external flooding event associated with the 23 ft still water flood (%EXTFL_1), this event is set to '0.0' in the model, while %EXTFL_2 is set to its nominal value. OPER-RESOSP and OPER-SWDISCHX are failed at both flooding scenarios, and therefore are set to 'TRUE' in the flag file. An error factor (EF) of 10 was added to each flooding initiating event to compensate for the uncertainty of their values. Quantification documentation was expanded to also include results and insights based on the cutset review.
| |
| | |
| BSEP 15-0101 Enclosure 3 Marked-up Technical Specification Pages - Unit 1
| |
| | |
| Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between the monitored Once within core keff and the predicted core keff is within + 1% Ak/k. 24 hours after reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 14 OONRI\I-MDT
| |
| '7 oprthecnse-dn MOE !~-H B
| |
| lIn accordance with the Surveillance Frequency Control Program
| |
| /
| |
| II Brunswick Unit 1 3.1-6 BrunwickUni 1 31-6Amendment No. 2-82
| |
| | |
| Control Rod OPERABILITY 3.1.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 2A. he'-rs L-.4 SR 3.1.3.2 ---------- NOTE-------.............
| |
| Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER greater than the L.PSP of the RWM.
| |
| ... is 4
| |
| 7' Insert each withdrawn control rod at least one notcl; m
| |
| SR 3.1.3.3 Verify each control rod scram time from fully In accordance with to notch position 06 is < 7 seconds. SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
| |
| In accordance with the Surveillance Frequency Control Program IJ Brunswick Unit 1 3.1-10 Bruswik Uit 3.-10Amendment No. 2-50
| |
| | |
| Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify, for a representative sample, each tested ,200-4...
| |
| control rod scram time is within the limits of c-wmIat-ive Table 3.1.4-1 with reactor steam dome pressure ..... t""n ""
| |
| > 800 psig. MOat SR 3.1.4.3 Verify each affected control rod scram time is wi in Prior to declaring the limits of Table 3.1.4-1 with any ratrsecontrol rod dome pressure. OPERABLE after work on control rod or CRD System that
| |
| * could affect scram time SR 3.1.4.4 Verify each affected control rod sc am time is within Prior to exceeding the limits of Table 3.1.4-1 with r trsemdome 40% RTP after fuel presure pig.movement__ 00 within the affected core cell in acordncewiththeAND Control Program Prior to exceeding 40% RTP after work on control rod or CR0 System that could affect scram time
| |
| ........*...L.Unit 1 i.*l *.lu u.,,,vv n*,u*
| |
| 3.1-13 Bfuswik No. 2-38 I Uit 3.-13Amendment
| |
| | |
| Control Rod Scram Accumulators 3.1.5 ACTIONS (continued) _________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control rod C.1 Verify all control rods Immediately upon scram accumulators associated with inoperable discovery of inoperable with reactor accumulators are fully charging water steam dome pressure inserted, header pressure
| |
| < 950 psig. < 940 psig AND C.2 Declare the associated 1 hour control rod inoperable.
| |
| D. Required Action B.1 or C.1 D.1-------NOTE- --
| |
| and associated Completion Not applicable if all Time not met. inoperable control rod scram accumulators are associated with fully inserted control rods.
| |
| Manually scram the reactor. Immediately SURVEILLANCEREQUIREMENTS _______
| |
| SU RVEI LLANCE FREQ UENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure is 7-d*y
| |
| Ž 940 psig. '
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3,1-17 BrunwickUnit13.-17Amendment No.
| |
| | |
| Rod Pattern Control 3.1.6 ACTIONS (continued)_________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1------------NOTE-----
| |
| control rods not in Control rod may be compliance with BPWS. bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1.
| |
| Suspend withdraw'al of Immediately control rods.
| |
| AND B.2 Manually scram the reactor. 1 hour SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE I FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with BPWS. hc'-r~i~s SIn accordance with the Surveillance Frequency Control Program I tl Brunswick Unit 1 3.1-19 BrunwickUnitI3.-19Amendment No. 24O*
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.I Restore SLC subsystem to 7 days inoperable. OPERABLE status.
| |
| B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable, subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate solution 24-hc'-rs is within the limits of Figure 3.1.7-1,.,
| |
| (continued)
| |
| Brunswick Unit 1 3.1-20 Amendment No. 2O2,
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate solution is 24-hetrs within the limits of Figure 3.1.7-2.
| |
| * SR 3.1.7.3 Verify temperature of pump suction and discharge/ 2A-heur-s SR 3.1.7.4 Verify continuity of explosive charge.
| |
| * SR 3.1.7.5 Verify the concentration of boron in so-uio1-witysL the limits of Figure 3.1.7-1.*,E' AND Once within In wth accodance the24 hours afterwae Contrl Prgramsolution temperature is restored within the limits of Figure 3.1.7-2 (continued)
| |
| Brunswick Unit 1 3.1-21 Bruswck ni I .121Amendment No. 22-7
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate Ž_41.2 gpm at a In accordance with discharge pressure > 1190 psig. the Inservice Testing Program SR 3.1.7.7 Verify flow through one SLC subsystem from pump 24-mont,,s-on
| |
| * into reactor pressure vessel. STAGGERED SR 3.1.7.8 Verify sodium pentaborate enrichment is > 47 ao Prior to addition to percent B-10. SLC tank I
| |
| IIn accordance with the Surveillance Frequency Control Program r
| |
| I Brunswick Unit 1 3.1-22 Bruswck ni I .122Amendment No. 227
| |
| | |
| SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1---------------------NOTE---------------
| |
| Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.
| |
| Verify each SDV vent and drain valve is open. 2A4-eays SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully 3!-days closed and fully open position.//
| |
| SR 3.1.8.3 Verify each SDV vent and drain valve: 2",cth
| |
| : a. Closes in < 30 seconds after receipt of a ra *ua or simulated scram signal; and V
| |
| : b. Opens when the actual or simulated s r -*
| |
| signal is reset. /
| |
| f In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.1-26 Brunsick Uit I .1-26Amendment No. O
| |
| | |
| APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER _>23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not A.1 Restore APLHGR(s) to 4 hours within limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS S URVEI LLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND 21 hou-r thereafter IIn accordance with the Surveillance Frequency Control Program jI II II Brunswick Unit 1 3.2-1 BrunwickUni I 32-1Amendment No. 2 MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| LCO 3.2.2 specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER _>23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A,.1 Restore MCPR(s) to within 4 hours limits, limits.
| |
| B. Required Action and 53.1 Reduce THERMAL 4 hours asoitdCompletion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified in the COLR. *12 hours after
| |
| Ž23% RTP AND 212, hourc thercaftcr (continued)
| |
| Brunswick Unit 1 3.2-2 Amendment No. 2-2 I
| |
| | |
| LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER >Ž23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž 23% RTP AND In accordance with the ~I Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.2-4 Bruswik Uit 3.-4Amendment No. 24
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES---------------
| |
| : 1. Refer to Table 3.3.1.1-1Ito determine which SRs apply for each RPS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.1 (Not used.)
| |
| SR 3.3.1.1.2 Perform CHANNEL CHECK. 24-heurs SR 3.3.1.1.3----------NOTE----------
| |
| Not required to be performed until 12 hours after THF:RMAI PO)WFR > 2)3o~/, IPTP Adjust the average power range monitor (APRM) channels to conform to the calculated power while operating at >_23% RTP.
| |
| SR 3.3.1.1.4 ------------- NOTE-------
| |
| Not required to be performed when entering MODI from MODE 1 until 12 hours after entering MODEi Perform CHANNEL FUNCTIONAL TEST.
| |
| '1 (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-4 BrunwickUni I 33-4Amendment No. 2 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform a functional test of each automatic scram 7dy contactor.
| |
| SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to withdrawing intermediate range monitor (IRM) channels overlap. SRMs from the fully Only bemet rquire dringsetryeitopMOD2tfro to MODE 1.
| |
| Verify the IRM and APRM channels overl* . -ay SR 3.3.1.1.8 Calibrate the local power range monit s.200cetifl I
| |
| SR 3.3.1.1.9 Perform CHANNEL FUNCTIONA TEST. 92dy SR 3.3.1.1.10 Calibrate the trip units. /_/_2-days...
| |
| I (continued)
| |
| In accordance with the Surveillance Frequency Control Program III I1*
| |
| Brunswick Unit 1 3.3-5 Bruswik Uit 3.-5Amendment No. 2-84
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.11 -~~~~NOTES----- --
| |
| : 1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
| |
| : 2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.
| |
| Perform CHANNEL FUNCTIONAL TEST. A 4-84-deys SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. S24 months SR 3.3.1.1.13 1. -~~~NOTES---
| |
| Neutron detectors are excluded.'- --
| |
| : 2. For Function 1, not required to be perf when entering MODE 2 from MODE 1 12 hours after entering MODE 2.
| |
| : 3. For Functions 2.b and 2.f, the recir transmitters that feed the APRMs Perform CHANNEL CALIBRATION. S2"!,months SR 3.3.1.1.14 (Not used.)
| |
| SR 3.3.1.1.15 Perform LOGIC SYSTEM FUN(
| |
| 2,A- months (continued)
| |
| In accordanceFrequency with the Surveillance Control Program l
| |
| I l
| |
| Brunswick Unit 1 3.3-6 Brunsick Uit 1 .3-6Amendment No. 24-I-
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve--Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure--Low Functions are not bypassed when THERMAL POWER '4 is >_26% RTP.
| |
| I SR 3.3.1.1.17 -~~~~NOTES-------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. For Functions 3 and 4, the sensor response timqi may be assumed to be the design sensor !
| |
| response time./
| |
| For Function 5, "n" equals I channels for the BASIS Frcquency.
| |
| For Function 2.e, "n" equals 8 channels for h D A C'IO I-'*
| |
| r'3*r.] If *+;*n I:
| |
| outputI., 4. sh,,.ll a'lte'rnate.
| |
| =
| |
| Verify the RPS RESPONSE TIME is within Ii 21! months on a STAGGERED
| |
| =
| |
| SR 3.3.1.1.18 Adjust recirculation drive flow to conform, core flow./ afer reaching equilibrium
| |
| ~'hce within 7 days conditions following refueling outage (continued)
| |
| In accordance with the Surveillance Frequency Control Program d
| |
| Brunswick Unit 1 3.3-7 Bruswik Uit 3.-7Amendment No. 2 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM Simulated 2A-,mciths Thermal Power is _>25% and recirculation drive flow is.,
| |
| < 60%. z In accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.3-8 BrunwickUni I 33-8Amendment No. 24-7
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS
| |
| -------- NOTE------------------------------...............
| |
| Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK.!2h-r SR 3.3.1.2.2----------NOTES---------
| |
| : 1. Only required to be met during CORE ALTERATIONS.
| |
| : 2. One SRM may be used to satisfy more than on*
| |
| of the following.
| |
| Verify an OPERABLE SRM detector is located in: 424hours
| |
| : a. The fueled region;,,
| |
| : b. The core quadrant where CORE ALTERATI N are being performed, when the associated R is included in the fueled region; and II (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-14 BrunwickUnitI No. 24-7 I 3.-14Amendment
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.4 ........................NOTES--------
| |
| : 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
| |
| : 2. Not required to be met during a core spiral offload.
| |
| Verify count rate is __3.0 cps.
| |
| ALTERATIONS AN4P 24-hc'-rs SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.1.2.6 ----------- NOTE Not required to be performed until 12 hoursj on Range 2 or below./
| |
| Perform CHANNEL FUNCTIONAL TES ! deys (continued)
| |
| L Surveillance Frequency In accordance Control with the Program Brunswick Unit 1 3.3-t5 Bruswik Uit No. 24-7 I 3.-15Amendment
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.7----------NOTES---------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
| |
| Perform CHANNEL CALIBRATION. 24-months I
| |
| Surveillance Frequency Control Program Brunswick Unit 1 3.3-16 No. 24 I BrunwickUnitI 3.-16Amendment
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor Mode E.1 Suspend control rod Immediately Switch--Shutdown Position withdrawal.
| |
| channels inoperable.
| |
| AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing onie or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
| |
| : 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
| |
| SURVEILLANCE/FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. .-1-84-1ay-s I (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 1 3.3-20 BruswikUit 3.-20Amendment No. 24-
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.2-----------NOTE- -------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn at < 8.75% RTP in MODE 2.
| |
| Perform CHANNEL FUNCTIONAL TEST. *'2-4a...
| |
| SR 3.3.2.1.3-------------------NOTE----------------
| |
| Not required to be performed until 1 hour after THERMAL POWER is *<8.75% RTP in MODE 1.
| |
| Perform CHANNEL FUNCTIONAL TEST. a 4..
| |
| SR 3.3.2.1.4 Verify the RBM: 2-~~h
| |
| : a. Low Power Range--Upscale Function 0 Intermediate Power Range--Upscale F ci OR High Power Range--Upscale Func o enabled (not bypassed) when APRM S ae Thermal Power is >_29%.
| |
| : b. Intermediate Power Range--Upscal Fnci OR High Power Range--Upscale F ini enabled (not bypassed) when APRI' mle Thermal Power is > Intermediate PF Rg Setpoint specified in the COLR.
| |
| : c. High Power Range--Upscale Fu i i enabled (not bypassed) when A imulated Thermal Power is _>High Poe e Setpoint specified in the COLR.
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-21 BrunwickUnitI 3.-2 1Amendment No. 22-2
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when THERMAL 24 mcnt~he POWER is
| |
| * 8.75% RTP.
| |
| SR 3.3.2.1.6-- - - - -- - - - -NOTE-.........................
| |
| Not required to be performed until 1 hour after reactorj mode switch is in the shutdown position. /
| |
| Perform CHANNEL FUNCTIONAL TEST. 24 mRths SR 3.3.2.1.7 -------- NOTE------------......
| |
| Neutron detectors are excluded.
| |
| Perform CHANNEL CALIBRATION. 2A-Renthe SR 3.3.2.1.8 Verify control rod sequences input to the conformance with BPWS. RWM OPERABLE folowing loading of
| |
| *ePIior to declaring squence into RM SIn accordance with the Surveillance Frequency Control Program II l
| |
| II Brunswick Unit 1 3.3-22 Bruswik Uit 3.-22Amendment No. 22-2
| |
| | |
| Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE REQUIREMENTS
| |
| ---------------- I'J I--
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided feedwater and main turbine high water level trip capability is maintained.
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL. CHECK. 2A-hour s aJ SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The Allowable A-months Value shall be < 207 inches.
| |
| SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, including valve actuation.
| |
| SIn accordanceFrequency Surveillance with the Control Program l
| |
| I m
| |
| Brunswick Unit 1 3.3-25 BrunwickUnitI 3.-25Amendment No. 2Oa
| |
| | |
| PAM Instrumentation 3.3.3.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| : 0. Required Action and 0.1 Enter the Condition Immediately associated Completion Time referenced in of Condition C not met. Table 3.3.3.1-1 for the channel.
| |
| E. As required by Required E.1 Be in MODE 3. 12 hours Action D.1 and referenced in Table 3.3.3.1-1.
| |
| F. As required by Required F.1 Initiate action in accordance Immediately Action D.1 and referenced in with Specification 5.6.6.
| |
| Table 3.3.3.1-1.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----------------
| |
| These SRs apply to each Function in Table 3.3.3.1-1.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK.3--dy Brunswick Unit 1 3.3-27 Amendment No. 2-34
| |
| | |
| PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.3 Perform CHANNEL CALIBRATION for each required 24-m,,c~ths PAM Instrumentation channel *_, I In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-28 Bruswck ni 1 .328Amendment No. 2-34
| |
| | |
| Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Remote Shutdown Monitoring Instrumentation LCO 3.3.3.2 The Remote Shutdown Monitoring Instrumentation Functions shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS
| |
| ------------- NC)
| |
| Separate Condition entry is allowed for each Function. I CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable, to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| S URVEI LLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required ,3- ,d...
| |
| instrumentation channel that is normally energized. ,,
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| /
| |
| Brunswick Unit 1 3.3-30 Bruswik Uit 3.-30Amendment No. 2-3-3
| |
| | |
| Remote Shutdown Monitoring Instrumentation 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each required 24 m,,cths instrumentation channel. =
| |
| SIn accordance with the Surveillance Frequency Control Program II l Brunswick Unit 1 3.3-31 Brunsick Uit I .3-31Amendment No. O
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One Function with B.1 Restore ATWS-RPT trip 72 hours ATWS-RPT trip capability capability.
| |
| not maintained.
| |
| C. Both Functions with C.1 Restore ATWS-RPT trip 1 hour ATWS-RPT trip capability capability for one Function.
| |
| not maintained.
| |
| D. Required Action and D.1 Remove the associated 6 hours associated Completion Time recirculation pump(s) from not met. service.
| |
| O__R D.2 Be in MODE 2. 6 hours SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----- ----------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 24-heut-s (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-33 Brunsick Uit I .3-33Amendment No. 2~
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.4.1.3 SR 3.3.4.1.4 Calibrate the trip units.
| |
| Perform CHANNEL CALIBRATION. The Allowable Q249 Values shall be:
| |
| : a. Reactor Vessel Water Level--Low Level 2:
| |
| >_101 inches; and
| |
| : b. Reactor Vessel Pressure--High: *<1147 psit SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST including breaker actuation. 24 meqths In accordance with the Surveillance Frequency Control Program d
| |
| Brunswick Unit 1 3.3-34 BrunwickUnitI 3.-34Amendment No. 2-a
| |
| | |
| ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS
| |
| ...................................... NOTES- - - - - - - - - - - - - - -
| |
| : 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c; and (b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
| |
| Brunswick Unit 1 3.3-40 Bruswik Uit 3.-40Amendment No. 2-§2
| |
| | |
| RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.
| |
| Brunswick Unit 1 3.3-47 Brunsick Uit I .3-47Amendment No. O
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES---------------
| |
| : 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions 2.c, 2.d, 3.a, 3.b, 3.e, 3.f, 3.g, 3.h, 4.a, 4.b, 4.e, 4.f, 4.g, 4.h, 4.i, 4.k, 5.a, 5.b, 5.e, 5.f, and 6.a; and (b) for up to 6 hours for all other Functions provided the associated Function maintains isolation capability.
| |
| Brunswick Unit 1 3.3-52 BrunwickUnitI No. 2-g$
| |
| 3.-52Amendment
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SU RVEI LLANCE IFREQ UENCY SR 3.3.6.1.8 ------------ NOTES---------
| |
| : 1. Radiation detectors are excluded.
| |
| : 2. The sensor response time for Functions l .a and 1.c may be assumed to be the design sensor response time.
| |
| Verify the ISOLATION INSTRUMENTATION RESPONSE TIME is within limits. ESTG C,,EIE SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST.
| |
| I In accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.3-53 Brunsick Uit I .3-53Amendment No. Q
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS ____________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 hour secondary containment isolation dampers inoperable.
| |
| AND C.2.1 Place the associated 1 hour standby gas treatment (SGT) subsystem(s) in operation.
| |
| O__R C.2.2 Declare associated SGT 1 hour subsystem(s) inoperable.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 24. hers (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program yI Brunswick Unit 1 3.3-60 Brunsick Uit I .3-60Amendment No. O
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92-dtays SR 3.3.6.2.3 Calibrate the trip unit. 2-In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-61 Brunswck Unt I 33-6 1Amendment No. O
| |
| | |
| CREV System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREV initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 24-.he'-rs SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. Q2dy SR 3.3.7.1.3 Perform CHANNEL CALIBRATION. , .A-.morths Brunswick Unit 1 3.3-64 Bruswck ni 1 .364Amendment No. 2<3
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Isolate condenser vacuum 12 hours associated Completion Time pumps.
| |
| of Condition A not met.
| |
| O__R OR B.2 Isolate main steam lines. 12 hours Condenser vacuum pump isolation capability not O.RR maintained.
| |
| B.3 Be in MODE 3. 12 hours SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation capability.
| |
| SU RVEI LLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK. a4-h9ur-s SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST.
| |
| ,Q2-onth SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The A Value shall be < 6 x background. /#
| |
| (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program I
| |
| Brunswick Unit 1 3.3-67 Bruswck ni I .367Amendment No. 2-*
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 2A-memthe including condenser vacuum pump trip breaker and isolation valve actuation.
| |
| IIn accordance with the Surveillance Frequency Control Program II I Brunswick Unit 1 3.3-68 Bruswck ni I .368Amendment No. 2-3
| |
| | |
| LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS
| |
| ..............--....................... NOTES- - - - - - - - - - - - - - -
| |
| : 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Functions maintains initiation capability for three DGs; and (b) for Function 2, the associated Function maintains DG initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST. 3-1-,4...
| |
| SR erformCHANNE.CALIBATION.48-meth 3.38.1.2 SR 3.3.8.1.3 Perform CHANNEL CALIBRATION. 18 -inths SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TES . 2A-mcFths IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-70 Bruswik Uit 3.-70Amendment No. 23Q
| |
| | |
| RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to fully insert Immediately associated Completion Time all insertable control rods in of Condition A or B not met core cells containing one or in MODE 3, 4, or 5 with any more fuel assemblies.
| |
| control rod withdrawn from a core cell containing one or more fuel assemblies.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.1 ------------ NOTE- -------
| |
| Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 for
| |
| _>24 hours.
| |
| Perform CHANNEL FUNCTIONAL TEST. 4~4-d~s SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS 2A. meRh:
| |
| motor generator set electric power monitoring assembly. The Allowable Values shall be:
| |
| : a. Overvoltage < 129 V.
| |
| : b. Undervoltage >_105 V.
| |
| : c. Underfrequency _>57.2 Hz.
| |
| (continued) lIn accordance with the II Surveillance Frequency Control Program If I
| |
| Brunswick Unit 1 3.3-73 BrunwickUnit1 3.-73Amendment No.* 2 I
| |
| | |
| RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS 2-4months alternate power supply electric power monitoring assembly. The Allowable Values shall be:
| |
| i,
| |
| : a. Overvoltage _<132 V.
| |
| : b. Undervoltage > 108 V.
| |
| : c. Underfrequency >_57.2 Hz.
| |
| SR 3.3.8.2.4 Perform a system functional test. I24-ment4h I
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-74 Bruswck ni 1 .374Amendment No. 2398
| |
| | |
| Recirculation Loops Operating 3.4.1 ACTIONS (continued) ________________
| |
| COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met.
| |
| O__R No recirculation loops in operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1------------NOTE---- -----
| |
| Not required to be performed until 24 hours after both recirculation loops are in operation.
| |
| Verify recirculation loop jet pump flow mismatch with 24-hetr-s both recirculation loops in operation:
| |
| : a. < 10% of rated core flow when operating at
| |
| < 75% of rated core flow; and
| |
| : b. < 5% of rated core flow when operating at /
| |
| _>75% of rated core flow.
| |
| In accordance with the Surveillance Frequency Control Program
| |
| * .-.---.-II Brunswick Unit 1 3.4-2 BrunwickUni I 34-2Amendment No. 2-44
| |
| | |
| Jet Pumps 3.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1-----------NOTES- -------
| |
| : 1. Not required to be performed until 4 hours after associated recirculation loop is in operation.
| |
| : 2. Not required to be performed until > 25% RTP.
| |
| Verify at least one of the following criteria (a or b) is 2-4-heu*s satisfied for each operating recirculation loop: *
| |
| : a. Recirculation pump flow to speed ratio differs by
| |
| <_5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by < 5% from established patterns.
| |
| : b. Each jet pump diffuser to lower plenum differential pressure differs by < 10% from that jet pump's established pattern.
| |
| In accordance with the l Surveillance Frequency Control Program Brunswick Unit 1 3.4-4 BrunwickUni I 34-4Amendment No. 24
| |
| | |
| SRVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.4.3.2-----------NOTE---.......................
| |
| Not required to be performed until 12 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify each required SRV opens when manually 2"-FPmonthS actuated.
| |
| IIn accordance with the Surveillance Frequency Control Program II l Brunswick Unit 1 3.4-6 Bruswik Uit 3.-6Amendment No. 2-2
| |
| | |
| RCS Operational LEAKAGE 3.4.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Pressure boundary LEAKAGE exists.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and 8-hei-s unidentified LEAKAGE increase are within limits. ,,
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.4-8 Brunsick Uit I .4-8Amendment No. O
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required primary 12-hours containment atmosphere radioactivity monitoring system.
| |
| SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of require*I leakage detection instrumentation. /,4 d~ys SR 3.4.5.3 .Perform a CHANNEL CALIBRATION of required leakage detection instrumentation.
| |
| Surveillance Frequency IIn accordance with the Control Program Ib I
| |
| Brunswick Unit 1 3.4-11 Bruswik Uit No. 2-g2 3.-11Amendment
| |
| | |
| RCS Specific Activity 3.4.6 ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours AND B.2.2.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1-------------NOTE--------
| |
| Only required to be performed in MODE 1.
| |
| Verify reactor coolant DOSE EQUIVALENT I-131 specific activity is < 0.2 IpCi/gm.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.4-13 Bruswik Uit 3.-13Amendment No. 2:Qg
| |
| | |
| RHR Shutdown Cooling System--Hot Shutdown 3.4.7 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.7.1-----------NOTE---------
| |
| Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR shutdown cooling isolation pressure.
| |
| Verify one required RHR shutdown cooling subsystem 42-heur-s or recirculation pump is operating.
| |
| In accordance with the Surveillance Frequency Control Program v/
| |
| q Brunswick Unit 1 3.4-16 Brunsick Uit 1 .4-16Amendment No. O
| |
| | |
| RHR Shutdown Cooling System--Cold Shutdown 3.4.8 ACTIONS (continued)_________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hour from subsystem in operation, circulating by an alternate discovery of no method. reactor coolant AND circulation No recirculation pump in AND operation.
| |
| Ornce per 12 hours thereafter AND B.2 Monitor reactor coolant Once per hour temperature.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling subsystem 4-2--heus or recirculation pump is operating.
| |
| SIn accordance with the II Surveillance Frequency Control Program p/
| |
| l Brunswick Unit 1 3.4-18 Brunsick Uit 1 .4-18Amendment No. Q
| |
| | |
| RCS P/T Limits 3.4.9 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C.------NOTE--...........C.1 Initiate action to restore Immediately Required Action 0.2 shall be parameter(s) to within completed if this Condition is limits.
| |
| entered.
| |
| -------- AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in'other than acceptable for operation. MODE 2 or 3.
| |
| MODES 1, 2, and 3.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.1---------------------NOTE---------------
| |
| Only required to be performed during RCS heatup and cooldown operations.
| |
| Verify: 3-mn,*if,, *,,
| |
| : a. RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2; and
| |
| : b. RCS heatup and cooldown rates are < 100°F in any 1 hour period.
| |
| (continued)
| |
| Surveillance Frequency Control Program I In accordance with the Brunswick Unit 1 3.4-20 Brunsick Uit I .4-20Amendment No. 0
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY
| |
| +
| |
| SR 3.4.9.2 - -- - - - - NOTE.................
| |
| Only required to be performed during RCS inservice leak and hydrostatic testing.
| |
| Verify: 24-Riates
| |
| : a. RCS pressure and RCS temperature are within the applicable limits specified in Figure 3.4.9-3; 3.4.9-4, or 3.4.9-5, as applicable.
| |
| : b. RCS heatup and cooldown rates are < 30°F in any 1 hour period.
| |
| SR 3.4.9.3 Verify RCS pressure and RCS temperature are within nce within the criticality limits specified in Figure 3.4.9-2. i5 minutes prior to ov"ntrol rod ithdrawal for the purpose of achieving criticality SIR 3.4.9.4-----------NOTE-----------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the bottom head cool ntOnce within temperature and the reactor pressure vessel (RPV 30 minutes prior to coolant temperature is < 145°F. each startup of a S recirculation pump I]
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program m m Brunswick Unit 1 3.4-21 Uit 3.-21Amendment Bruswik No. 2-2-8
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS ('continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.5-----------NOTE----------------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the reactor coolant Once within temperature in~the recirculation loop to be started and 30 minutes prior, to the RPV coolant temperature is < 50°F. each startup of a recirculation pump SR 3.4.9.6-----------NOTE----------------
| |
| Only required to be performed when tensioning the reactor vessel head bolting studs.
| |
| Verify reactor vessel flange and head flange 8-iue temperatures are > 70°F.
| |
| SR 3.4.9.7-----------NOTE---------
| |
| Not required to be performed until 30 minutes after RCS temperature _<80°F in MODE 4.
| |
| Verify reactor vessel flange and head flange ~O-R~*Jtes temperatures are > 70°F.
| |
| I SR 3.4.9.8 ----------- NOTE--------
| |
| Not required to be performed until 12 hours after R(
| |
| temperature < 100°F in MODE 4.
| |
| Verify reactor vessel flange and head flange temperatures are _>70°F. 42-eJ
| |
| "" S*
| |
| Y In accordance with the Surveillance Frequency Control Program l
| |
| I dl Brunswick Unit 1 3.4-22 BrunwickUnitI3.-22Amendment No. 2-g8
| |
| | |
| Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be _<1045 psig.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is < 1045 psig. !2--he'-s I
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.4-28 BrunwickUnitI No. 22-8 I 3.-28Amendment
| |
| | |
| ECCS--Operating 3.5.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME J. Two or more low pressure J.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A or B.
| |
| OR HPCI System and two or more required ADS valves inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, the 34dy piping is filled with water from the pump discharge valve to the injection valve.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program t
| |
| II II Brunswick Unit 1 3.5-4 BrunwickUni I 35-4Amendment No. 2Oa-
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.2------------NOTE--------
| |
| Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling isolation pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each ECCS injection/spray subsystem manual, ~-1-,-d...
| |
| power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in /
| |
| position, is in the correct position.
| |
| SR 3.5.1.3 Verify ADS pneumatic supply header pressure is '*-1,-d...
| |
| > 95 psig.
| |
| SR 3.5.1.4 Verify the RHR System cross tie valve is locked 3 4..
| |
| closed.
| |
| SR 3.5.1.5------------NOTE--------
| |
| Not required to be performed if performed withi t previous 31 days.
| |
| Verify each recirculation pump discharge val a Once each startup bypass valve cycles through one complete leffl rior to exceeding travel or is de-energized in the closed posi 25% RTP (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program l
| |
| I II Brunswick Unit 1 3.5-5 Brunsick Uit I No. O
| |
| .5-5Amendment
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.6 Verify the following ECCS pumps develop the 92~de~,LS specified flow rate against a system head corresponding to the specified reactor pressure. #4 SYSTEM HEAD CORRESPONDI NC NO. OF TO A REACTOR.
| |
| SYSTEM FLOW RATE PUMPS PRESSURE OF I CS __4100gpm 1 ___113 psig LPCi > 14,000 gpm 2 >_20 psig SR 3.5.1.7 ------------ NOTE- -----
| |
| Not required to be performed until 48 hours aft*.
| |
| reactor steam pressure is adequate to perforrrl test. I Verify, withpump reactor Q2-daas the HPCI unitpressure < 1045 can develop and rat a flow Ž>*45 psig,
| |
| > 4250 gpm against a system head corre* onding tc reactor pressure./
| |
| SR 3.5.1.8 Not required to be performed until 48 ours reactor----------
| |
| steam pressure is NOTE adequate t / perfor test. I 2A-memths Verify, with reactor pressure < I1 pump unit can develop a flow rat a system head corresponding to (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.5-6 BrunwickUni I 35-6Amendment No. 24
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.9------------NOTE--------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each ECCS injection/spray subsystem actuates 2"!.meithe on an actual or simulated automatic initiation signal.
| |
| * SR3511-----------OTE-----------
| |
| Valve actuation may be excluded.
| |
| Verify the ADS actuates on an actual or simulated__2"A.menths automatic initiation signal. 1 SR 3.5.1.11-----------NOTE---------
| |
| Not required to be performed until 12 hours aft*
| |
| reactor steam pressure is adequate to perfor th test.
| |
| Verify each required ADS valve opens whe mauly2A-meih Verify the ECCS RE sPOS wTiME oreahCOS 24-meat-h injection/spray subsystemiswtn lmtz SSurveillance IIn accordanceFrequency with the Control Program Brunswick Unit 1 3.5-7 BrunwickUni I 35-7Amendment No.-2O*
| |
| | |
| ECCS-Shutdown 3.5.2 ACTIONS (continued) ________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 and D.1 Initiate action to restore Immediately associated Completion Time secondary containment to not met. OPERABLE status.
| |
| AND D.2 Initiate action to restore one Immediately standby gas treatment subsystem to OPERABLE status.
| |
| AND D.3 Initiate action to restore Immediately isolation capability in each required secondary containment penetration flow path not isolated.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure coolant injection !2-hc'-s (LPCI) subsystem, the suppression pool water level is In witacoranc th t*(continued)
| |
| Brunswick Unit 1 3.5-9 Amendment No. Q
| |
| | |
| ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 4
| |
| SR 3.5.2.2 Verify, for each required core spray (CS) subsystem, the:
| |
| : a. Suppression pool water level is >_-31 inches; or b.---------NOTE--- ---------- /
| |
| Only one required CS subsystem may take credit for this option during OPDRVs./
| |
| SR 3.5.2.3 Condensate storage tank water volume is
| |
| >_228,200 gallons.
| |
| Verify, for each required ECCS injection/spray I 2-ty subsystem, the piping is filled with water from the A pump discharge valve to the injection valve. !
| |
| SR 3.5.2.4------------NOTE----------------.....
| |
| One LPCI subsystem may be considered OPE BL during alignment and operation for decay heatrmoI if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each required ECCS injection/spray su s enb-~
| |
| manual, power operated, and automatic valv ith flow path, that is not locked, sealed, or otheri secured in position, is in the correct position (continued)
| |
| In accordance with the Surveillance Frequency Control Program
| |
| * J Brunswick Unit 1 3.5-10 BrunwickUnitI 3.-10Amendment No. 2-Q8
| |
| | |
| ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 92deL SR 3.5.2.5 Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure.
| |
| SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS > 113 psig
| |
| __4100gpm 1 LPCI > 9000 gpm I > 20 psig SR 3.5.2.6 ------------- NOTE--------------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each required ECCS injection/spray subsysten 14Rmonths actuates on an actual or simulated automatic initiatioi signal.
| |
| S R 3 .5 .2.7 - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - -.--- - -
| |
| Instrumentation response time may be assumed to b/
| |
| the design instrumentation response time.
| |
| Verify the ECCS RESPONSE TIME for each require*
| |
| ECCS injection/spray subsystem is within the limit. 12A-meat-hs In accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 1 3.5-11 BrunwickUnitINo. 2§2, 3.-11Amendment
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water from -- de the pump discharge valve to the injection valve. 1 SR 3.5.3.2 Verify each RCIC System manual, power operated, 2.1-day and automatic valve in the flow path, that is not locke sealed, or otherwise secur~ed in position, is in the correct position.
| |
| SR 3.5.3.3------------NOTE--...................
| |
| : 1. Use of auxiliary steam for the performan o the SR is not allowed.
| |
| : 2. Not required to be performed until 24 o sate reactor steam pressure is adequate t prom the test.
| |
| Verify, with reactor pressure _Ž 945 psig d< 1045 -dS psig, the
| |
| Ž__400 gpmRCIC pump against can develop a system head acor*
| |
| fib ponditng ate to , "
| |
| reactor pressure. °7 (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.5-13 Bruswik Uit 3.-13Amendment No. 2-82
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.4 -~~~~NOTES--------
| |
| : 1. Use of auxiliary steam for the performance of the SR is not allowed with reactor pressure
| |
| > 150 psig.
| |
| : 2. Not required to be performed until 24 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify, with turbine inlet pressure Ž_135 psig and < 165 2A-Renthe psig, the RCIC pump can develop a flow rate
| |
| _Ž400 gpm against a system head corresponding to an.
| |
| equivalent reactor pressure.
| |
| I SR 3.5.3.5 -~~~~NOTE--------.
| |
| Vessel injection may be excluded. It Verify the RCIC System actuates on an actual or simulated automatic initiation signal.
| |
| IIn accordance with the Surveillance Frequency Control Program I
| |
| I II Brunswick Unit 1 3.5-14 Brunsick Uit I .5-14Amendment No. O
| |
| | |
| Primary Containment 3.6.1.1 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and leakage rate In accordance with testing, except for primary containment air lock testing, the Primary in accordance with the Primary Containment Leakage Containment Rate Testing Program. Leakage Rate Testing Program SR 3.6.1.1 .2 Verify drywell to suppression chamber differential 4-ment-hs pressure does not decrease at a rate > 0.25 inch water iA gauge per minute tested over a 10 minute period at an initial differential pressure ofŽ> 1.00 psid and
| |
| * 1.25 psid.*-
| |
| I n accordance with the Surveillance Frequency Control Program t/
| |
| Brunswick Unit 1 3.6-2 Brunsick Uit I .6-2Amendment No. -
| |
| | |
| Primary Containment Air Lock 3.6.1.2 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.2.1-----------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 3.6.1.1.1.
| |
| Perform required primary containment air lock leakage In accordance with rate testing in accordance with the Primary the Primary Containment Leakage Rate Testing Program. Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary containment air / mcnteRhs lock can be opened at a time. I Inaccordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.6-6 Amendment No. 2O2,
| |
| | |
| PClVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 -...............
| |
| NOTES---------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment isolation manual S1-days valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. /
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program
| |
| ¢'
| |
| Brunswick Unit 1 3.6-11 Brunsick Uit I .6-11Amendment No. O
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.2 ------------ NOTES- -------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing incore probe (TIP) av, 34....~
| |
| shear isolation valve explosive charge. I SR 3.6.1.3.4 Verify the isolation time of each power operated and Inraccordance each automatic PCIV, except for MSIVs, is within with the Inservice limits. Testing Program SR 3.6.1.3.5 Verify the isolation time of each MSIV is _> 3 secotd In accordance with and *<5 seconds. the Inservice Testing Program Inaccordance with the (cniud Surveillance Frequency Control Program Brunswick Unit 1 3.6-12 Amendment No. 24*
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 1ieth position on an actual or simulated isolation signal.
| |
| SR 3.6.1.3.7 Verify a representative sample of reactor I2A-R~eRthe instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.
| |
| SR 3.6.1.3.8 Remove and test the explosive squib from each r'the n accordance isolation valve of the TIP System. Inservice with Testing Program SR 3.6.1.3.9 Verify leakage rate through each main ste*
| |
| < 100 scfh and the combined leakage rate the Primary main steam lines is _<150 scfh when testec. Containment SIn accordance with
| |
| > 25 psig. Leakage Rate Testing Program In accordance with the Surveillance Frequency Control Program 'i d
| |
| Brunswick Unit 1 3.6-13 Bruswck ni I .613Amendment No. 2-39
| |
| | |
| Drywell Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Air Temperature LCO 3.6.1.4 Drywell average air temperature shall be _<150°F.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell average air 8 hours temperature not within limit, temperature to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE/FREQUENCY SR 3.6.1.4.1 Verify drywell average air temperature is within limit. 24he*
| |
| IControl Program Brunswick Unit 1 3.6-14 Brunsick Uit I .6-14Amendment No. O
| |
| | |
| Reactor Building-to-Suppression Chamber Vaccum Breakers 3.6.1.5 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify nitrogen bottle supply pressure of each nitrogen 2A-hetrs backup subsystem is _>1130 psig. _
| |
| SR 3.6.1.5.2 ----------- NOTES--------
| |
| : 1. Not required to be met for vacuum breakerst are open during Surveillances./
| |
| : 2. Not required to be met for vacuum breaker/
| |
| open when performing their intended on Verify each vacuum breaker is closed.
| |
| SR 3.6.1.5.3 Perform a functional test of each vacuum_
| |
| SR 3.6.1.5.4 Verify the f 2A-R~eRth~
| |
| _<0.5 psid.
| |
| SR 3.6.1.5.5 Verify leakage rate of each nitroger subsystem is < 0.65 scfm when tes 24 e*. t4h e, nitrogen bottle supply pressure of j SR 3.6.1.5.6 Verify the Nitrogen Backup Sy I2A-meR~hs to the vacuum breakers on an actuation signal. j Surveillance Frequency In accordance with the Control Program I
| |
| Brunswick Unit 1 3.6-17 Brunsick Uit I .6-17Amendment No. 2~
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 -NOTE Not required to be met for vacuum breakers that are open during Surveillances.
| |
| Verify each vacuum breaker is closed.
| |
| Within 6 hours after any discharge of steam to the suppression chamber from any source AND Within 6 hours following an operation that causes any of the vacuum breakers to open SR 3.6.1.6.2 Perform a functional test of each rq §2-de~s breaker.
| |
| AND Within 12 hours after any SIn accordance with the discharge of Surveillance Frequency steam to the Control Program suppression II chamber from the SRVs SR 3.6.1.6.3 Verify the full open setpoint of each required vaum 2~4-imiths breaker is < 0.5 psid.
| |
| Brunswick Unit 1 3.6-19 BrunwickUnitI No. 2&-84 I 3.-19Amendment
| |
| | |
| Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature is within the applicable limits. A SAND 5 minutes when performing testing that adds heat to the suppression pool
| |
| /
| |
| In accordance with the Surveillance Frequency Control Program
| |
| /
| |
| Brunswick Unit 1 3.6-22 Brunsick Uit I .6-22Amendment No. O
| |
| | |
| Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be > -31 inches and _<-27 inches.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression pool 2 hours level not within limits, water level to within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits. 2-e*
| |
| SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.6-23 Brunsick Uit I .6-23Amendment No. O
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pooi cooling subsystem ~1d~
| |
| manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise /
| |
| while operating in the suppression pool cooling oe In accordance with the Surveillance Frequency Control Program iy I
| |
| Brunswick Unit 1 3.6-25 Brunsick Uit I .6-25Amendment No. O
| |
| | |
| Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LCO 3.6.3.1 The primary containment oxygen concentration shall be < 4.0 volume percent.
| |
| APPLICABILITY: MODE 1 during the time period:
| |
| : a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to'
| |
| : b. 24 hours prior to a scheduled reduction of THERMAL POWER to
| |
| < 15% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment oxygen A.1 Restore oxygen 24 hours concentration not within concentration to within limit.
| |
| limit.
| |
| B. Required Action and B.1 Reduce THERMAL 8 hours associated Completion Time POWER to _<15% RTP.
| |
| not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.'1.1 Verify primary containment oxygen concentration is 7-41ays within limits.
| |
| Surveillance Frequency Control Program Brunswick Unit 1 3.6-26 Amendment No. 243
| |
| | |
| Secondary Containment 3.6.4.1 ACTIONR£ COMPLETION CONDITION REQUIRED ACTION TIME C. (continued) C.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment equipment hatches 2A.-iciths are closed and sealed.
| |
| SR 3.6.4.1.2 Verify one secondary containment access door is ~A-meRths closed in each access opening.
| |
| SR 3.6.4.1.3 Verify each SGT subsystem can maintain > 0.25 incl S/rT'4,AGG ERED of vacuum water gauge in the secondary cnane for 1 hour at a flow rate<* 3000 cfm. I I
| |
| Iin accordance with the Surveillance Frequency Control Program l
| |
| * I Brunswick Unit 1 3.6-29 Bruswik Uit 3.-29Amendment No. 252 I
| |
| | |
| SCl~s 3.6.4.2 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME D. Required Action and D.1 -NOTE----
| |
| associated Completion Time LCO 3.0.3 is not applicable.
| |
| of Condition A or B not met - - - -
| |
| during movement of recently irradiated fuel assemblies in Suspend movement of Immediately the secondary containment recently irradiated fuel or during OPDRVs. assemblies in the secondary containment.
| |
| AND D.2 Initiate action to suspend Immediately OPDRVs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.2.1 Verify the isolation time of each automatic SCID is 2-eth within limits.
| |
| lI A
| |
| SR 3.6.4.2.2 Verify each automatic SCID actuates to the isolation position on an actual or simulated actuation signal.
| |
| In accordance with the Surveillance Frequency Control Program I
| |
| J Brunswick Unit 1 3.6-32 Bruswik Uit 3.-32Amendment No. 2-* I
| |
| | |
| SGT System 3.6.4.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two SGT subsystems E.1I-NOTE----.....
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of recently irradiated fuel - - -
| |
| assemblies in the secondary containment or during Suspend movement of Immediately OPDRVs. recently irradiated fuel assemblies in secondary containment.
| |
| AND E.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for _ 10 continuous "3 -d ...
| |
| hours with heaters operating.
| |
| SR 3.6.4.3.2 Perform required SGT filter testing in accordance In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.6.4.3.3 2A-m......hs Verify eachinitiation simulated signal. actuates on an actual c SGT subsystem IIn accordance with the Surveillance Frequency Control Program I
| |
| I I
| |
| Brunswick Unit 1 3.6-35 Bruswik Uit 3.-35Amendment No. 252-
| |
| | |
| RHRSW System 3.7.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION !COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual, power operated, and automatic valve in the flow path, that is not locked, I, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.
| |
| SIn accordance with the Surveillance Frequency Control Program U
| |
| Brunswick Unit 1 3.7-3 BrunwickUni I 37-3Amendment No. 2Oa-
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the SW pump suction bay of 2A-.hc,-r the intake structure is > -6 ft mean sea level. ,
| |
| SR 3.7.2.2 Verify the water temperature of UHS is _<90.5°F.
| |
| t-SR 3.7.2.3 -------- NOTE---------------........
| |
| Isolation of flow to individual components does not j render SW System inoperable. I Verify each SW System manual, power opera -4e~s automatic valve in the flow paths servicing sa related systems or components, that is not Io*
| |
| sealed, or otherwise secured in position, is in correct position.
| |
| SR 3.7.2.4 ----------- NOTES------
| |
| : 1. A single test at the specified Frequenc*
| |
| satisfy this Surveillance for both units.A
| |
| : 2. Isolation of flow to individual compor not render SW System inoperable.
| |
| Verify automatic transfer of each DG coo Q2-deaLs supply from the normal SW supply to th supply on low OG jacket cooling water pressure.I (continued)
| |
| In accordance with the Surveillance Frequency Control Program d
| |
| Brunswick Unit 1 3.7-9 Bruswik Uit 3.-9Amendment N o.-2-1-,
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.5-----------NOTE---------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each required SW System automatic component 2!.4-menth*
| |
| actuates on an actual or simulated initiation signal. ,,,
| |
| In accordance with the Surveillance Frequency Control Program p/
| |
| Brunswick Unit 1 3.7-10 BrunwickUnitI 3.-10Amendment No. 203
| |
| | |
| CREV System 3.7.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two CREV subsystems--------NOTE-------
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of irradiated fuel assemblies- ------------
| |
| in the secondary E.1 Suspend movement of Immediately containment, during CORE irradiated fuel assemblies ALTERATIONS, or during in the secondary OPDRVs. containment.
| |
| OR AND On rmrRVE.2 Suspend CORE Immediately subsystems inoperable due ATRTOS to an inoperable CRE boundary during movement AND of irradiated fuel assemblies E.3 Initiate action to suspend Immediately in the secondaryOPRs containment, during COREOPRs ALTERATIONS, or during OPD RVs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREV subsystem for _ 15 continuous 21d*
| |
| minutes.
| |
| SR 3..3.2 Perform required CREV filter testing in accordance* In accordance with SR.7..2 with the Ventilation Filter Testing Program (VFTP)./ the VFTP (continued)
| |
| Su nt re*l pnCg rFreq u e n cY I--t In accordance with the I /
| |
| Brunswick Unit 1 3.7-13 BruswikUit 3.-13Amendment No. 248
| |
| | |
| CREV System 3.7.3 SURVEILLANCE REQUI REMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.3 Perform required CRE unfiltered air inleakage testing In accordance with in accordance with the Control Room Envelope the Control Room Habitability Program. Envelope Habitability Program SR 3.7.3.4 Verify each CREV subsystem actuates on an actual or 24-moinths simulated initiation signal. .
| |
| In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.7-14 Bruswik Uit 3.-14Amendment No. 248
| |
| | |
| Control Room AC System 3.7.4 ACTIONS (continued)_________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME F. Three control room AC---------------NOTE---------
| |
| subsystems inoperable LCO 3.0.3 is not applicable.
| |
| during movement of irradiated fuel assemblies in the secondary containment, F.1 Suspend movement of Immediately during CORE irradiated fuel assemblies in ALTERATIONS, or during the secondary containment.
| |
| OP DRVs.
| |
| AND F.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND F.3 Initiate actions to suspend Immediately OPDR Vs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.'7.4.1 Verify each control room AC subsystem has the2",m th capability to remove the assumed heat load.
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.7-17 BrunwickUnitINo. 24O3 3.-17Amendment
| |
| | |
| Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4
| |
| SR 3.7.5.1 -~~~~~NOTE---------
| |
| Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
| |
| Verify the gross gamma activity rate of the noble 31 dacys gases is *<243,600 pCi/second otter decay ot AND 30 minutes.
| |
| Once within 4 hours after aŽ>50%
| |
| increase in the nominal steady state fission gas In accordance with the release after
| |
| ,Surveillance Frequency factoring out Control Program increases due to l changes in THERMAL POWER level Brunswick Unit 1 3.7-19 Bruswik Uit No. 2-O2 3.-19Amendment
| |
| | |
| Main Turbine Bypass System 3.7.6 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete cycle of each main turbine bypass 34-ey valve.
| |
| SR 3.7.6.2 Perform a system functional test. 2-iFh SR 3.7:6.3 Verify the TURBINE BYPASS SYSTEM RESPON TIME is within limits.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.7-21 BrunwickUnitI 3.-2 1Amendment No. 24*
| |
| | |
| Spent Fuel Storage Pool Water Level 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Spent Fuel Storage Pool Water Level LCO 3.7.7 The spent fuel storage pool water level shall be _ 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage racks.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.
| |
| ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool A.1------------NOTE------
| |
| water level not within limit. LCO 3.0.3 is not applicable.
| |
| Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the spent fuel storage pool water level is
| |
| _ 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage racks. /.
| |
| Brunswick Unit 1 3.7-22 Amendment No. O
| |
| | |
| AC Sources--Operating 3.8.1 SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each offsite circuit.
| |
| '4 SR 3.8.1.2
| |
| : 1. -------
| |
| All DG starts mayNOTES---------
| |
| prelube period.
| |
| be preceded by an engine
| |
| //
| |
| : 2. A modified DG start involving idling and gra, acceleration to synchronous speed may be for this SR. When modified start procedure:.
| |
| not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
| |
| : 3. A single test at the specified Frequency41 satisfy this Surveillance for both units.
| |
| Verify achieveseachsteady DG starts state from 4 standby conditio and voltage >Ž3750 V and_*4300 V and frequency _>58.8 Hz and <*61.2 Hz.
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.8-7 Bruswik Uit 3.-7Amendment No. 26
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.3 S~~~~NOTES--------
| |
| : 1. DG loadings may include gradual loading.
| |
| : 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.
| |
| : 5. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG is synchronized and loaded and 31-deys operates for _ 60 minutes at a load > 2800 kW and
| |
| _<3500 kW. "
| |
| SR 3.8.1.4 Verify each engine mounted tank contains >_150 fuel oil. I aldy SR 3.8.1.5 Check engine for and remove mounted tank. accumulated water fr SR 3.8.1.6 Verify the fuel oil transfer system orc fuel oil from the day fuel oil storage mounted tank.
| |
| (continued)
| |
| Surveillance Frequency SIn accordance Control Program with the Ib I
| |
| 4 Brunswick Unit 1 3.8-8 Bruswik No. 2-0 I Uit 3.-8Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.7 ------------ NOTES---------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| IA Verify each OG starts from standby condition and achieves, in < 10 seconds, voltage Ž_3750 V and frequency_> 58.8 Hz, and after steady state conditions are reached, maintains voltage _>3750 V and
| |
| * 4300 V and frequency > 58.8 Hz and _<61.2 Hz.
| |
| -. 5.
| |
| (continued) lIn accordance with the Surveillance Frequency Control Program
| |
| * i Brunswick Unit 1 3.8-9 BrunwickUni I No. 2-06 I 38-9Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.8 ----------- NOTES- -------
| |
| : 1. SR 3.8.1 .8.a shall not be performed in MODE 1 or 2 for the Unit 1 offsite circuits. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. SR 3.8.1.8.a is not required to be met if the unit power supply is from the preferred offsite circuit.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify:
| |
| : a. Automatic transfer capability of the unit power supply from the normal circuit to the preferred 2AReV h offsite circuit; and
| |
| : b. Manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsitei circuit. /
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program III Brunswick Unit 1 3.8-10 BrunwickUnitI 3.-10Amendment No. 2-g8
| |
| | |
| AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.9 -~~~NOTES--------
| |
| : 1. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. If performed with the DG. synchronized with offsite power, it shall be performed at a power factor < 0.9.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG rejects a load greater than or equal to its associated core spray pump without tripping. 2A- eth (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program 1 1 Brunswick Unit 1 3.8-11 Bruswik Uit 3.-11Amendment No. 2-
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.10 -~NOTE---------------
| |
| A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG's automatic trips are bypassed on an
| |
| /
| |
| actual or simulated ECCS initiation signal except:
| |
| : a. Engine overspeed;
| |
| : b. Generator differential overcurrent;
| |
| : c. Low lube oil pressure; 1
| |
| : d. Reverse power;
| |
| : e. Loss of field; and
| |
| : f. Phase overcurrent (voltage restrained).
| |
| (continued)
| |
| Control Program Brunswick Unit 1 3.8-12 BrunwickUnitI No. 2-88 I 3.-12Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.11 ----------- NOTES- -------
| |
| : 1. Momentary transients outside the load and power factor ranges do not invalidate this test.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG operating at a power factor _<0.9 operates for > 60 minutes loaded to >_3500 kW and
| |
| < 3850 kW. 4 A4m~4 SR 3.8.1.12 A single test at the specified Frequency will satisfy thi
| |
| -------- NOTE---------
| |
| Surveillance for both units.]
| |
| Verify an actual or simulated ECCS initiation signal~ls
| |
| ~4-R~eHthe capable of overriding the test mode feature to retuf each DG to ready-to-load operation.
| |
| (continued) i IIn accordanceFrequency Surveillance with the Control Program l
| |
| Brunswick Unit 1 3,8-13 Bruswik Uit 3.-13Amendment No.
| |
| * AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.13--------------------NOTE----------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 1 and OG 2. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify interval between each sequenced load block is 2A-,months within + 10% of design interval for each load sequence IJ relay.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program
| |
| * I Brunswick Unit 1 3.8-14 Bruswik Uit 3.-14Amendment No. 2- I
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.14-----------NOTES- -------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify, on actual or simulated loss of offsite power 2".mth signal initiationin conjunction signal: with an actual or simulated ECOS
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses; and
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in
| |
| < 10.5 seconds,
| |
| : 2. energizes auto-connected emergency loads through load sequence relays,
| |
| : 3. maintains steady state voltage _>3750 V and <*4300 V,
| |
| : 4. maintains steady state frequency > 58.8 Hz and
| |
| * 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected emergency loads for
| |
| __5 minutes.
| |
| SIn accordance with the I Surveillance Frequency Control Program
| |
| * II Brunswick Unit 1 3.8-15 BrunwickUnitI 3.-15Amendment No. 2 6
| |
| | |
| Diesel Fuel Oil 3.8.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.3.1 For each required DG, verify:v, .. *
| |
| : a. The associated day fuel oil storage tank contains Ž_22,650 gal; and
| |
| : b. The main fuel oil storage tank contains >_20,850 gal per required DG.
| |
| SR 3.8.3.2 Verify fuel oil properties of stored fuel oil are tested ii In accordance with accordance with, and maintained within the limits of the Diesel Fuel Oil the Diesel Fuel 0il Testing Program. / STesting Program I
| |
| SR 3.8.3.3 Check for and remove accumulated water from e*
| |
| day fuel oil tank and the main fuel oil storage tang m m SIn accordanceFrequency with the Surveillance Control Program
| |
| =
| |
| Brunswick Unit 1 3.8-22 BrunwickUnitI No. 2O 3.-22Amendment I
| |
| | |
| DC Sources--Operating 3.8.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Two or more DC electrical power subsystems inoperable.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is _>130 V on float 7-4a~&
| |
| charge.
| |
| /
| |
| SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.
| |
| O__R Verify battery connection and resistance is _<23.0 p ohn
| |
| _< 82.8 tpohms for int1 for inter-cell connections rack connections.I SR 3.8.4.3 Verify battery cells, cell plates, and racks show !8-Imenths visual indication of physical damage or abnormj deterioration that degrades performance. /
| |
| (continued)
| |
| SIn accordanceFrequency Surveillance with the Control Program fl I
| |
| I Brunswick Unit 1 3.8-24 BrunwickUnit1 No. 2O I 3.-24Amendment
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell !8-imcnths and terminal connections are coated with anti-corrosion material./
| |
| SR 3.8.4.5 Verify each required battery charger supplies /24-menths
| |
| Ž>250 amps at ___135 V for Ž>4 hours.im, SR 3.8.4.6 ----------- NOTES--------
| |
| : 1. The modified performance discharge test in SR 3.8.4.7 may be performed in lieu of the service test in SR 3.8.4.6 once per 60 months.
| |
| : 2. This Surveillance shall not be performed in MULJL 1 or 2 tor 1 UUL electrical pov thle unitcredit subsystems. However, may be taken unplanned events that satisfy this SR.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify battery capacity is adequate to supply, and I24 mcn~he maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.8-25 Brunsick Uit I .8-25Amendment No.20
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.7 -~~~~NOTES---------
| |
| : 1. This Surveillance shall not be performed in MODE 1 or 2 for the Unit I DC electrical power subsystems. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify battery capacity is _Ž80% of the manufacturer's *0 Rmenths rating when subjected to a performance discharge tesp or a modified performance discharge test. /
| |
| 12 months when battery shows degradation or has reached 85% of the expected life with IIn accordance with the capacity < 100% of ISurveillance Frequency
| |
| * manufacturer's IControl Program rating A._ND 24 months when battery has reached 85% of the expected life with capacity Ž_100% of manufacturer's rating L ___________________________
| |
| Brunswick Unit 1 3.8-26 BrunwickUnitI No. 205 I 3.-26Amendment
| |
| | |
| Battery Cell Parameters 3.8.6 ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.
| |
| B. Required Action and B.1 Declare associated battery 'Immediately
| |
| .associated Completion Time inoperable.
| |
| of Condition A not met.
| |
| OR One or more batteries with average electrolyte temperature of the representative cells not within limits.
| |
| OR One or more batteries with one or more battery cell parameters not within Category C limits.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 7-d4ay Category A limits.
| |
| (continued)
| |
| LSurveillance Frequency Inaccordance Control Program with the j
| |
| Brunswick Unit 1 3.8-31 Brunsick Uit No. O I I .8-31Amendment
| |
| | |
| Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 "2-a" Category B limits.
| |
| SR 3.8.6.3 Verify average electrolyte temperature of representative cells is >_60°F.
| |
| In accordanceFrequency with the Surveillance Control Program I
| |
| l Brunswick Unit 1 3.8-32 BrunwickUnitI No. 29 I 3.-32Amendment
| |
| | |
| Distribution Systems--Operating 3.8.7 SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and indicated power 7dy availability to required AC and DC electrical power ,
| |
| distribution subsystems.
| |
| SR 3.8.7.2 Verify no combination of more than two power 7dy conversion modules (consisting of either two lighi inverters or one lighting inverter and one plant uninterruptible power supply unit) are aligned to Division II bus B.
| |
| In accordance with the Surveillance Frequency Control Program
| |
| _//
| |
| Brunswick Unit 1 3.8-37 BrunwickUnitI 3.-37Amendment No. 2O I
| |
| | |
| Distribution Systems-Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to suspend Immediately operations with a potential for draining the reactor vessel.
| |
| AND
| |
| *A.2.4 Initiate actions to restore Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and indicated power 7-d*
| |
| availability to required AC and DC electrical power distribution subsystems. 7 In accordance with the Surveillance Frequency Control Program
| |
| _/
| |
| Brunswick Unit 1 3.8-39 Bruswik Uit 3.-39Amendment No. 2O6
| |
| | |
| Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of 7-deys the following required refueling equipment interlock inputs:
| |
| : a. All-rods-in,
| |
| : b. Refuel platform position,
| |
| : c. Refuel platform fuel grapple, fuel loaded,
| |
| : d. Fuel grapple position,
| |
| /
| |
| : e. Refuel platform frame-mounted hoist, fuel loaded, and
| |
| : f. Refuel platform monorail hoist, fuel loaded.
| |
| In accordance with the Surveillance Frequency Control Program 1z/
| |
| ,I Brunswick Unit 1 3.9-2 BrunwickUni I 39-2Amendment No. 202,
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
| |
| ACTIONS_________________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-out A.1 Suspend control rod Immediately interlock inoperable, withdrawal.
| |
| AND A.2 Initiate action to fully insert Immediately all insertable control rods in
| |
| *core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in Refuel position. !2-liurs (continued) lIn accordance with the Surveillance Frequency yU Control Program Brunswick Unit 1 3.9-3 Brunsick Uit I .9-3Amendment No. O
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| S URVE ILLANCE FREQUENCY SR 3.9.2.2-------------NOTE---------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn.
| |
| Perform CHANNEL FUNCTIONAL TEST. ~ dy A
| |
| I SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.9-4 Brunsick Uit I No. O
| |
| .9-4Amendment
| |
| | |
| Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
| |
| APPLICABILITY: When loading fuel assemblies into the core.
| |
| ACTIONS_________________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control rods A.1 Suspend loading fuel Immediately not fully inserted, assemblies into the core.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE JFREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted. 42-hcurc In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.9-5 Brunsick Uit I .9-5Amendment No. O
| |
| | |
| Control Rod OPERABILITY--Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY--Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.
| |
| APPLICABILITY: MODE 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to fully insert Immediately control rods inoperable, inoperable withdrawn control rods.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1-------------NOTE----------------........
| |
| Not required to be performed until 7 days after the control rod is withdrawn.
| |
| Insert each withdrawn control rod at least one notch. ~-d4ays SR 3.9.5.2 Verify each withdrawn control rod scram accumulato/r pressure is > 940 psig. /f V IIn accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.9-8 Brunsick Uit I No. o
| |
| .9-8Amendment
| |
| | |
| RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 RPV water level shall be _> 23 ft above the top of irradiated fuel assemblies seated within the RPV.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies within the RPV, During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.
| |
| ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not within A.1 Suspend movement of fuel Immediately limit. assemblies and handling of control rods within the RPV.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is >_23 ft above the top of 2'!.he-rs irradiated fuel assemblies seated within the RPV. t I
| |
| In accordance with the Surveillance Frequency Control Program t_/
| |
| I Brunswick Unit 1 3.9-9 BrunsickUit I .9-9Amendment No.
| |
| | |
| RHR-High Water Level 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is 42-hc'-r,,
| |
| operating. j SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.9-12 Bruswik Uit 3.-12Amendment No. 2-
| |
| | |
| RHR--Low Water Level 3.9.8 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is !-2-hc,-r operating.
| |
| Control Program Brunswick Unit 1 3.9-15 Brunsick Uit I .9-15Amendment No. 2~
| |
| | |
| Reactor Mode Switch Interlock Testing 3.10.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Place the reactor mode 1 hour switch in the shutdown position.
| |
| OR A.3.2------NOTE----
| |
| Only applicable in MODE 5.
| |
| Place the reactor mode 1 hour switch in the refuel position.
| |
| SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted in core cells 42-he~f S containing one or more fuel assemblies.
| |
| SR 3.10.2.2 Verify no CORE ALTERATIONS are in 2A-.ho'-s I
| |
| In accordance with the Surveillance Frequency Control Program m
| |
| Brunswick Unit 1 3,10-5 BrunwickUnitI 3.0-5Amendment No. 2-Q8
| |
| | |
| Single Controi Rod Withdrawal--Hot Shutdown 3.10.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required LCOs. According to the Sapplicable SRs SR 3.10.3.2 ----------- NOTE- --------
| |
| Not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d. 1 requirements.
| |
| Verify all control rods, other than the control rod being 24-heur-s withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| SR 3.10.3.3 Verify all control rods, other than the control rod be 2A4-heu e withdrawn, are fully inserted.
| |
| lIn accordance with the Surveillance Frequency Control Program l
| |
| I II Brunswick Unit 1 3.10-8 Brunswck Unt I 310-8Amendment No. 2~
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 ACTIONS (continued)__________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One or more of the above B.1 Suspend withdrawal of the Immediately requirements not met with control rod and removal of the affected control rod not associated CRD.
| |
| insertable.
| |
| AND B.2.1 Initiate action to fully insert Immediately all control rods.
| |
| O__R B.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.4.2----------NOTE- --------
| |
| Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c. 1 requirements.
| |
| Verify all control rods, other than the control rod being 24-heI4F withdrawn, in a five by five array centered on the ,
| |
| control rod being withdrawn, are disarmed. 1
| |
| _/
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.10-11 BrunwickUnitI 3.0-11Amendment No. 2Q*
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS SR 3.10.4.3 Verify all control rods, other than the control rod being 24-heuws withdrawn, are fully inserted.
| |
| I. -
| |
| W ;
| |
| SR 3.10.4.4 Not required to be met if SR 3.10.4.1 is satisfiedfo/
| |
| LCO 3.10.4.b.1
| |
| -- -- -- OTE-----------Io requirements./
| |
| Verify a control rod withdrawal block is inserted. 24-heur-s In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.10-12 Brunwic Unt I3.1-12Amendment No. 24§
| |
| | |
| Single CR0 Removal--Refueling 3.10.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Initiate action to fully insert Immediately all control rods.
| |
| OR A.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods, other than the control rod 24-heurs withdrawn for the removal of the associated CR0, arei1t fully inserted.
| |
| SR 3.10.5.2 Verify all control rods, other than the control rod 21-.hc'-rs withdrawn for the removal of the associated CR0, i ,a five by five array centered on the control rod withd awn for the removal of the associated CR0, are disarred SR 3.10.5.3 Verify a control rod withdrawal block is inserte 2A-heius SR 3.10.5.4 Perform SR 3.1.1.1..**kcrigt (continued)
| |
| In accordance with the Surveillance Frequency Control Program F-f,,
| |
| l Brunswick Unit 1 3.10-14 Brunwic Unt I3.1-14Amendment No. 282
| |
| | |
| Single CRD Removal--Refueling 3.10.5 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.5 Verify no other CORE ALTERATIONS are in progress. 2A-heur-s I
| |
| SIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.10-15 Brunwic Unt I3.1-15Amendment No. 2O2,
| |
| | |
| Multiple Control Rod Withdrawal--Refueling 3.10.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Initiate action to fully insert Immediately all control rods in core cells containing one or more fuel assemblies.
| |
| O__R A.3.2' Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.6.1 Verify the four fuel assemblies are removed from core 2!hus cells associated with each control rod or CRD removed.
| |
| SR 3.10.6.2 Verify all other control rods in core cells containing 01 2A, hours or more fuel assemblies are fully inserted.
| |
| SR 3.10.6.3 ----------- NOTE-------
| |
| Only required to be met during fuel loading. ,
| |
| Verify fuel assemblies being loaded are in ,2A-.he'ds with an approved spiral reload sequence.
| |
| IIn accordanceFrequency Surveillance with the Control Program l
| |
| II Brunswick Unit 1 3.10-17 Brunwic Unt I3.1-17Amendment No. 2-00
| |
| | |
| SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.2----------NOTE- --------
| |
| Not required to be met if SR 3.10.8.3 satisfied.
| |
| Perform the MODE 2 applicable SRs for LCO 3.3.2.1, According to the Function 2 of Table 3.3.2.1-1. applicable SRs SR 3.10.8.3----------NOTE- --------
| |
| Not required to be met if SR 3.10.8.2 satisfied.
| |
| Verify movement of control rods is in compliance with During control rod the approved control rod sequence for the SDM test by movement a second licensed operator or other qualified member of the technical staff.
| |
| SR 3.10.8.4 Verify no other CORE ALTERATIONS are in progress. 12-hc'-rs
| |
| -j (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.10-22 Brunswck Unt1 310-22Amendment No. O
| |
| | |
| SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.5 Verify each withdrawn control rod does not go to the -Eachtime the withdrawn overtravel position. control rod is withdrawn to "full out" position
| |
| _AND Prior to satisfying LCO 3.10.8.c requirement after work on control rod or CRD System that could affect coupling SR 3.10.8.6 Verify CRD charging water header pressure > 940 psig.
| |
| IIn accordance with the II Surveillance Frequency Control Program S
| |
| Brunswick Unit 1 3.10-23 Brunswck UntI 310-23Amendment No.20
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals Control Room Envelope Habitability Procqram (continued)
| |
| : e. The quantitative limits on unfiltered air inleakage into the ORE. 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 ORE 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 ORE habitability, determining ORE unfiltered inleakage, and measuring ORE pressure and assessing the ORE boundary as required by paragraphs c and d, respectively.
| |
| 41=
| |
| 5.5.14 Surveillance Frequency Control Prociram This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| : 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. Ohanges 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.
| |
| Brunswick Unit 1 5.0-17a BrunwickUnit1No. 248 I 5.-17aAmendment
| |
| | |
| BSEP 15-0101 Enclosure 4 Marked-up Technical Specification Pages - Unit 2
| |
| | |
| Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between the monitored Once within core keff and the predicted core keff is within+/-+1% Ak/k. 24 hours after reaching equilibrium conditions following startup after fuel mnovement within the reactor. pressure vessel or control rod replacement AND 14,!00MWD7 ni MODE !
| |
| I naccordance with the Surveillance Frequency ljI Control Program Brunswick Unit 2 3.1-6 Brunsick Uit 2 .1-6Amendment No. 2
| |
| | |
| Control Rod OPERABILITY 3.1.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 2.4--heurs SR 3.1.3.2 ---------- NOTE- --------
| |
| Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
| |
| 7' Insert each withdrawn control rod at least one notch 3m~
| |
| SR 3.1.3.3 Verify each control rod scram time from fully n accordance with to notch position 06 is < 7 seconds. SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.1-10 Bruswik No. 2-78 I Uit 3.-10Amendment
| |
| | |
| Controi Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify, for a representative sample, each tested control ,200-da...
| |
| rod scram time is within the limits of Table 3.1.4-1 with eumu*lati-ve reactor steam dome pressure _>800 psig. "p....÷;"- ;t SR 3.1.4.3 Verify each affected control rod scram time is within nrror to declaring the limits of Table 3.1.4-1 with any reactor steam dom* control rod pressure. OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram tim* is within Prior to exceeding the limits of Table 3.1.4-1 with reactor st am dome 40% RTP after fuel pressure > 800 psig. /movement within the affected core cell AND In accordance with the I*-
| |
| Surveillance Frequency Prior to exceeding Control Program 40% RTP after work on control rod or CRD System that could affect scram time Brunswick Unit 2 3.1-13 Bruswik Uit No. 264' I 3.-13Amendment
| |
| | |
| Control Rod Scram Accumulators 3.1.5 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control rod C.1 Verify all control rods Immediately upon scram accumulators associated with inoperable discovery of inoperable with reactor accumulators are fully charging water steam dome pressure inserted, header pressure
| |
| < 950 psig. < 940 psig AND C.2 Declare the associated 1 hour control rod inoperable.
| |
| D. Required Action B.1 or C.1 D.1-------NOTE----
| |
| and associated Completion Not applicable if all Time not met. inoperable control rod scram accumulators are associated with fully inserted control rods.
| |
| Manually scram the reactor. Immediately SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure is 7-d4ays
| |
| _>940 psig.,j I n accordance with the Surveillance Frequency Control Program II l Brunswick Unit 2 3.1-17 Bruswik Uit 3.-17Amendment No. 2-32
| |
| | |
| Rod Pattern Control 3.1.6 ACTIONS (continued) ________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1-------NOTE----
| |
| control rods not in Control rod may be compliance with BPWS. bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1.
| |
| Suspend withdrawal, of Immediately control rods.
| |
| AND B.2 Manually scram the reactor. 1 hour SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with BPWS. 2-e~
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.1-19 BrunwickUnit2 3.-19Amendment No. 223
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS ___________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem to 7 days inoperable. OPERABLE status.
| |
| B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable, subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate solution 24-heir-s is within the limits of Figure 3.1.7-1. I,,*
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.1-20 Bruswik Uit 3.-20Amendment No. 2438
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate solution is 24-heuf within the limits of Figure 3.1.7-2.
| |
| SR 3.1.7.3 Verify temperature of pump suction and discharge I2A-h~s piping up to the SLC injection valves is withinthl of Figure 3.1.7-2. 1/h SR 3.1.7.4 Verify continuit SR 3.1.7.5 Verify the concentration of boron in solutic 4*yt the limits of Figure 3.1.7-1.
| |
| Once within In accordance with the
| |
| * 24 hours after water Surveillance Frequency or boron is added to Control Program solution I
| |
| AND Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-2 (continued)
| |
| Brunswick Unit 2 3.1-21 BrunwickUnit2 3.-21Amendment No. 25 I
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate _ 41.2 gpm at a In accordance with discharge pressure Ž>1190 psig the Inservice Testing Program SR 3.1.7.7 Verify flow through one SLC subsystem from pump 21! monthc on into reactor pressure vessel STAGGERED TEST SR 3.1.7.8 Verify sodium pentaborate enrichment is > 47 atom n*ror to addition to percent B-10. SLC tank I
| |
| IIn accordance with the Surveillance Frequency Control Program f
| |
| I Brunswick Unit 2 3.1-22 Bruswik Uit 3.-22Amendment No. 2-*5
| |
| | |
| SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 -NO TE-........................
| |
| Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.
| |
| Verify each SDV vent and drain valve is open.
| |
| A SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully closed and fully open position.
| |
| m SR 3.1.8.3 Verify each SDV vent and drain valve:
| |
| 2A-months
| |
| : a. Closes in < 30 seconds after receipt of ai or simulated scram signal; and
| |
| : b. Opens when the actual or simulated signal is reset.
| |
| SIn accordanceFrequency Surveillance with the Control Program i
| |
| I Ii Brunswick Unit 2 3.1-26 Bruswik Uit 3.-26Amendment No. 2-a2
| |
| | |
| APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| LCO 3.2.1 COLR.
| |
| APPLICABILITY: THERMAL POWER Ž_23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 4 hours limits. within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion Time POWER to < 23% RTP.
| |
| not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND 2hors hrcfc iF In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 2 3.2-1 Bruswik Uit 3.-1Amendment No. 24-7-i
| |
| | |
| MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| LCO 3.2.2 specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER Ž>23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 4 hours Iimits. limits.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND 24 hours thcreafer (continued)
| |
| In accordance with the Surveillance Frequency Control Program j'I Brunswick Unit 2 3.2-2 BrunwickUni 2 I 32-2Amendment No. 247-
| |
| | |
| LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER >Ž23% RTP.
| |
| ACTIONS_________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND t-her-eaf te SIn accordance with the Surveillance Frequency Control Program j'I l III Brunswick Unit 2 3.2-4 Bruswik Uit 3.-4Amendment No. 2--74 I
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS
| |
| ...................................... NOTES-----------------------------..............
| |
| : 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.1 (Not used.)
| |
| SR 3.3.1.1.2 Perform CHANNEL CHECK. 24-h~eur-SR 3.3.1.1.3 ------------ NOTE----------------
| |
| Not required to be performed until 12 hours after THERMAL POWER >Ž23% RTP.
| |
| \
| |
| Adjust the average power range monitor (APRM) channels to conform to the calculated power while hi operating at > 23% RTP.
| |
| SR 3.3.1.1.4-----------NOTE--------
| |
| Not required to be performed when entering MODE2, fr~om _MO_ DE_ 1_until 12 _h~ours_ after_ entering_ MOD E_ 2 ._
| |
| Perform CHANNEL FUNCTIONAL TEST.
| |
| 1l (continued)
| |
| In accordanceFrequency Surveillance with the Control Program II Brunswick Unit 2 3.3-4 Brunwic Uni 233-4Amendment No. 24-7
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform a functional test of each automatic scram i-day contactor.
| |
| SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to withdrawing intermediate range monitor (IRM) channels overlap SRMs from the fully inserted position SR 3.3.1.1.7--------------------.......NOTE--------
| |
| Only required to be met during entry into MOI2 ro MODEl1.
| |
| Verify the IRM and APRM channels overl p.
| |
| SR 3.3.1.1.8 Calibrate the local pwranemonit rs. 2000 ,cf,,ti,,cfull power rnge .. .............
| |
| SR 3.3.1.1.9 Perform CHANNEL FUNCTION ET §o ,d...
| |
| SR 3.3.1.1.10 Calibrate the trip units. Q112- ....
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 2 3.3-5 Bruswik Uit 3.-5AmendmentNo. 2-82 I
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.11 -~~~~NOTES----- ---
| |
| : 1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
| |
| : 2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.
| |
| Perform CHANNEL FUNCTIONAL TEST.
| |
| 444dt SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.1.1.13 -~~NOTES--------- I
| |
| : 1. Neutron detectors are excluded. )
| |
| : 2. For Function 1, not required to be perf when entering MODE 2 from MODE 1 12 hours after entering MODE 2. )
| |
| : 3. For Functions 2.b and 2.f, the rec transmitters that feed the APRMs Perform CHANNEL CALIBRATION. 24, mcnthe SR 3.3.1.1.14 (Not used.)
| |
| SR 3.3.1.1.15 Perform LOGIC SYSTEM FL LA -nenth*
| |
| (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program 1
| |
| I Brunswick Unit 2 3.3-6 BrunwickUni 2 33-6Amendment No. 242,I
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)__________
| |
| SURVEILLANCE JFREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve--Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure--Low Functions are not bypassed when THERMAL POWER is _>26% RTP.
| |
| SR 3.3.1.1.17 ---------------......... NOTES- ------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. For Functions 3 and 4, the sensor response /
| |
| (
| |
| time may be assumed to be the design senso-response time./
| |
| i* l," I*lii.-I#,*,l.*Al,*
| |
| * fl*fl * .,*1* A *1.. .... I* ,,il*_ J.I V.11111tz~I§ iui 1 r ,,q!Iue ,n,.e, 4 ........
| |
| ulluri .*I(:
| |
| 1" I TEcOT rA') "Ie - . ...
| |
| I-jul I rIIrlr'r"' ~*"~-"*
| |
| outputs ,- hll-*I a-lt+r-.atc..-
| |
| STAGGERED TEST Verify the RPS RESPONSE TIME is Il SR 3.3.1.1.18 Adjust the flow control trip refe* Once within 7 days to reactor flow. after reaching equilibrium conditions following refueling outage (continued)
| |
| SIn accordanceFrequency with the l Surveillance
| |
| * Control Programl II I Brunswick Unit 2 3.3-7 Bruswik Uit 3.-7Amendment No. 24-7
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM Simulated 24-meiaths Thermal Power is > 25% and recirculation drive flow is
| |
| < 60%.
| |
| In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.3-8 BrunwickUni 2 33-8Amendment No. 24*
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS
| |
| ----------------...................... NOTE----------------------------..............
| |
| Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. 12-hc,-rs SR 3.3.1.2.2 1.
| |
| 2.
| |
| -------- NOTES---------..............
| |
| Only required to be met during CORE ALTERATIONS.
| |
| One SRM may be used to satisfy more than of the following.
| |
| 1 Verify an OPERABLE SRM detector is located in:
| |
| : a. The fueled region; 42-hcurs
| |
| : b. The core quadrant where CORE ALTERA are being performed, when the associated is included in the fueled region; and /
| |
| : c. A core quadrant adjacent to where COF*
| |
| ALTERATIONS are being performed, v4h associated SRM is included in the fuel4!d SR 3.3.1.2.3 Perform CHANNEL CHECK. . a-eu-I (continued)
| |
| Im
| |
| * Surveillance Frequency lIn accordance Control Program with the Brunswick Unit 2 3.3-14 BrunwickUnit2 No. 24 I 3.-14Amendment
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.4 -~~~~NOTES--------
| |
| : 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
| |
| : 2. Not required to be met during a core spiral off load.
| |
| Verify count rate is >_3.0 cps.
| |
| AI4P-DATI*%
| |
| SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST.
| |
| lay SR 3.3.1.2.6 -~~NOTE----
| |
| Not required to be performed until 12 hours a*
| |
| on Range 2 or below. /
| |
| Perform CHANNEL FUNCTIONAL TES]
| |
| (continued Surveillance Frequency IIn accordance Control with the Program I
| |
| Brunswick Unit 2 3.3-15 Bruswik Uit 3.-15Amendment No. 242,
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.7------------NOTES--------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
| |
| Perform CHANNEL CALIBRATION. 24-m~em'ths Inaccordance with the SureilaceFrequency Control Program Brunswick Unit 2 3.3-16 BrunwickUnit2No. 24 I 3.-16Amendment
| |
| | |
| Control Rod Block Instrumentation 3.3.2,1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor Mode E.1 Suspend control rod Immediately Switch--Shutdown Position withdrawal.
| |
| channels inoperable.
| |
| AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ...................................... NOTES- - - - - - - - - - - - - - -
| |
| : 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
| |
| : 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
| |
| SURVEILLANCE FREUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. -- dy I (continued)
| |
| In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 2 3.3-20 Bruswik Uit 3.-20Amendment No.
| |
| * Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.2-----------NOTE- -------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn at _<8.75% RTP in MODE 2.
| |
| Not required to be performed until 1 hour after THERMAL POWER is*_<8.75% RTP in MODE 1.
| |
| Perform CHANNEL FUNCTIONAL TEST. -dy SR 3.3.2.1.4 Verify the RBM: 2A-ReRths
| |
| : a. Low Power Range--Upscale Function OR Intermediate Power Range--Upscale Fur ti OR High Power Range--Upscale Functic enabled (not bypassed) when APRM Si u e Thermal Power is _>29%.
| |
| : b. Intermediate Power Range--Upscale tio OR High Power Range--Upscale Fun t ni enabled (not bypassed) when APRM iua Thermal Power is _>Intermediate Po eRa Setpoint specified in the COLR.
| |
| : c. High Power Range--Upscale Func i enabled (not bypassed) when APRS lae Thermal Power is __High Power R* g etpoint specified in the COLR.
| |
| Y (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.3-21 BrunwickUnit2 No. 24-7 I 3.-21Amendment
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when THERMAL 2A-meat-s POWER is
| |
| * 8.75% RTP.
| |
| ;SH 3.3.Z. 1.f5 ------------- "U I -------------------.......
| |
| Not required to be performed until 1 hour after reactor*
| |
| mode switch is in the shutdown position.
| |
| Perform CHANNEL FUNCTIONAL TEST. 124-mGeth SR 3.3.2.1.7 ----------- NOTE-----------
| |
| Neutron detectors are excluded.
| |
| Perform CHANNEL CALIBRATION. 24-meiMhs SR 3.3.2.1.8 Verify control rod sequences input to the Prior to declaring conformance with BPWS. RWM OPERABLE following loading of sequence into RWM IIn accordanceFrequency Surveillance with the Control Program I
| |
| Brunswick Unit 2 3.3-22 BrunwickUnit2 3.-22Amendment No. 2-4 Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE------ ---------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided feedwater and main turbine high water level trip capability is maintained.
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK. 24-heus SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The Allowable 2me*
| |
| Value shall be < 207 inches. .
| |
| SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24, mcnth*
| |
| including valve actuation. ,1 In accordance with the Surveillance Frequency Control Program
| |
| !/
| |
| Brunswick Unit 2 3.3-25 BrunwickUnit2 3.-25Amendment No. 22*
| |
| | |
| PAM Instrumentation 3.3.3.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion Time referenced in of Condition C not met. Table 3.3.3.1-1 for the channel.
| |
| E. As required by Required E.1 Be in MODE 3. 12 hours Action D.1 and referenced in Table 3.3.3.1-1.
| |
| F. As required by Required F.1 Initiate action in accordance Immediately Action D.1 and referenced in with Specification 5.6.6.
| |
| Table 3.3.3.1-1.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE------ ---------
| |
| These SRs apply to each Function in Table 3.3.3.1-1.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. ,3-,-d....
| |
| SR 3.3.3.1.2 (Not Used.) /
| |
| I (continued)
| |
| In accordance with the Surveillance Frequency Control Program F/
| |
| ..m Brunswick Unit 2 3.3-27 Bruswck ni 2 .327Amendment No. 2 PAM Instrumentation 3.3.3.1 SURVEILLANCE REQU IREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.3 Perform CHANNEL CALIBRATION for each required 24-me~tthe PAM Instrumentation channel. .A I IIn accordance with the Surveillance Frequency Control Program l I Brunswick Unit 2 3.3-28 Bruswck ni 2 .328Amendment No. 26 Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Remote Shutdown Monitoring Instrumentation LCO 3.3.3.2 OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS*
| |
| ------------- NOTE Separate Condition entry is allowed for each Functih I
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable, to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 1dy instrumentation channel that is normally energized.
| |
| (continued) lIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 2 3.3-30 BrunwickUnit2 3.-30Amendment No. 2-60
| |
| | |
| Remote Shutdown Monitoring Instrumentation 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each required 2A-e*hs instrumentation channel.
| |
| In accordance with the Surveillance Frequency Control Program tl Brunswick Unit 2 3.3-31 Brunsick Uit 2 .3-31Amendment No. 2
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One Function with B.1 Restore ATWS-RPT trip 72 hours ATWS-RPT trip capability capability.
| |
| not maintained.
| |
| C. Both Functions with C.1 Restore ATWS-RPT trip 1 hour ATWS-RPT trip capability capability for one Function.
| |
| not maintained.
| |
| D. Required Action and D.1 Remove the associated 6 hours associated Completion Time recirculation pump(s) from not met. service.
| |
| O__R D.2 Be in MODE 2. 6 hours SURVEILLANCE REQUIREMENTS
| |
| ------------------------- I,.1 I--------------------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 2A. he'-r (continued)
| |
| IIn accordance with the qi Surveillance Frequency Control Program I II Brunswick Unit 2 3.3-33 BrunwickUnit2 3.-33Amendment No. 22*
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 1~
| |
| SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST. 92-days f-SR 3.3.4.1.3 Calibrate the trip units.
| |
| SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The Allowable a4-me~ths Values shall be:
| |
| : a. Reactor Vessel Water Level--Low Level *.
| |
| Ž_101 inches; and/
| |
| : b. Reactor Vessel Pressure--High: _<114 SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL. [2A-msiths including breaker actuation.
| |
| lIn accordanceFrequency Surveillance with the Control Program I
| |
| I I
| |
| Brunswick Unit 2 3.3-34 BrunwickUnit2 3.-34Amendment No. 22-a
| |
| | |
| ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS
| |
| ---------------...................... NOTES------ ---------
| |
| : 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c; and (b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.3-40 Brunsick Uit 2 .3-40Amendment No. 2
| |
| | |
| RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.
| |
| Brunswick Unit 2 3.3-47 Brunsick Uit 2 .3-47Amendment No. 2
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES---------------
| |
| : 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions 2.c, 2.d, 3.a, 3.b, 3.e, 3.f, 3.g, 3.h, 4.a, 4.b, 4.e, 4.f, 4.g, 4.h, 4.i, 4.k, 5.a, 5.b, 5.e, 5.f, and 6.a; and (b) for up to 6 hours for all other Functions provided the associated Function maintains isolation capability.
| |
| Brunswick Unit 2 3.3-52 Brunsick Uit 2 .3-52Amendment No. 2-
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.1.8 -~~~~NOTES----- --
| |
| : 1. Radiation detectors are excluded.
| |
| : 2. The sensor response time for Functions 1 .a, 1 .c, and 1.f may be assumed to be the design sensor response time.
| |
| Verify the ISOLATION INSTRUMENTATION RESPONSE TIME is within limits. ESTAGGERED SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST. i llm9 I~l i In accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 2 3.3-53 Brunsick Uit 2 .3-53Amendment No. 2
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 hour secondary containment isolation dampers inoperable.
| |
| AND C.2.1 Place the associated 1 hour standby gas treatment (SGT) subsystem(s) in operation.
| |
| O__R C.2.2 Declare associated SGT 1 hour subsystem(s) inoperable.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability.
| |
| SURVEILLANCE .LFREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 2-ew In accordance with the I J cniud Surveillance Frequency Control Program Brunswick Unit 2 3.3-60 Amendment No. 2,34
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 Brunswick Unit 2 3.3-61 BrunwickUnit2 3.-61Amendment No. 2-*
| |
| | |
| CREV System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREV initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. -- Q2-days SR 3.3.7.1.3 Perform CHANNEL CALIBRATION. 2A-meait-h Brunswick Unit 2 3.3-64 Bruswck ni 2 .364Amendment No. 267-
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Isolate condenser vacuum 12 hours associated Completion Time pumps.
| |
| of Condition A not met.
| |
| OR O__R B.2 Isolate main steam lines. 12 hours Condenser vacuum pump isolation capability not O__RR maintained.
| |
| B.3 Be in MODE 3. 12 hours SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK. 24-heur-s SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST. ~ye SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The Allc !48-months Value shall be _< 6 x background. /
| |
| (continued)
| |
| SIn accordanceFrequency Surveillance with the Control Program I
| |
| lII Brunswick Unit 2 3.3-67 Bruswck ni 2 .367Amendment No. 267-
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 24-meaths including condenser vacuum pump trip breaker and isolation valve actuation.-/
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.3-68 Bruswck ni 2 .368Amendment No. 2-6 LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS
| |
| --------- NOTES------------------------------..............
| |
| : 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Functions maintains initiation capability for three DGs; and (b) for Function 2, the associated Function maintains OG initiation capability.
| |
| Brunswick Unit 2 3.3-70 Bruswck ni 2 .370Amendment No. 267
| |
| | |
| RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to fully insert Immediately associated Completion Time all insertable control rods in of Condition A or B not met core cells containing one or in MODE 3, 4, or 5 with any more fuel assemblies.
| |
| control rod withdrawn from a core cell containing one or more fuel assemblies.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.1 ------------ NOTE---------
| |
| Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 for
| |
| >_24 hours.
| |
| Perform CHANNEL FUNCTIONAL TEST. 48A-deys SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS _A-mePths motor generator set electric power monitoring assembly. The Allowable Values shall be:
| |
| : a. Overvoltage _<129 V.
| |
| : b. Undervoltage > 105 V.
| |
| : c. Underfrequency > 57.2 Hz.
| |
| (continued)
| |
| In accordance with the v/
| |
| Surveillance Frequency Control Program II Brunswick Unit 2 3.3-73 Bruswck ni 2 .373Amendment No. 26 RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS 24-meat-hs alternate power supply electric power monitoring assembly. The Allowable Values shall be:
| |
| : a. Overvoltage _<132 V.
| |
| : b. Undervoltage _>108 V.
| |
| : c. Underfrequency _> 57.2 Hz.
| |
| SR 3.3.8.2.4 Perform a system functional test. 24-menths SIn accordance with the Surveillance Frequency Control Program
| |
| !II Brunswick Unit 2 3.3-74 Bruswck ni 2 .374Amendment No. 267-
| |
| | |
| Recirculation Loops Operating 3.4.1 ACTIONS (continued) _________
| |
| COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met.
| |
| OR No recirculation loops in operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1-----------NOTE- -------
| |
| Not required to be performed until 24 hours after both recirculation loops are in operation Verify recirculation loop jet pump flow mismatch with 2"4.heurs both recirculation loops in operation: ,
| |
| : a. < 10% of rated core flow when operating at
| |
| < 75% of rated core flow; and/
| |
| : b. < 5% of rated core flow when operating at A
| |
| _ 75% of rated core flow.
| |
| SIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.4-2 BrunwickUni2 34-2Amendment No. 2 Jet Pumps 3.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 -~~~~NOTES---------
| |
| : 1. Not required to be performed until 4 hours after associated recirculation loop is in operation.
| |
| : 2. Not required to be performed until _>25% RTP.
| |
| Verify at least one of the following criteria (a or b) is satisfied for each operating recirculation loop:
| |
| 2A1k
| |
| : a. Recirculation pump flow to speed ratio differs by
| |
| _<5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by _<5% from established patterns.
| |
| : b. Each jet pump diffuser to lower plenum differential pressure differs by _<10% from that jet pump's established pattern.
| |
| H-IIn accordance with the Surveillance Frequency Control Program l l Brunswick Unit 2 3.4-4 Bruswik Uit 3.-4Amendment No. 227
| |
| | |
| S RVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.3.2-----------NOTE----................
| |
| Not required to be performed until 12 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify each re~quired SRV opens when manually2",m th actuated.
| |
| In accordance with the Surveillance Frequency Control Program i
| |
| Brunswick Unit 2 3.4-6 Brunsick Uit 2 .4-6Amendment No.23
| |
| | |
| RCS Operational LEAKAGE 3.4.4 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Pressure boundary LEAKAGE exists.
| |
| SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and 8hu-unidentified LEAKAGE increase are within limits. .,
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 2 3.4-8 Brunsick Uit 2 No. ~
| |
| .4-8Amendment
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required primary --- hr-containment atmosphere radioactivity monitoring system.
| |
| SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of requi leakage detection instrumentation.
| |
| SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required 2A-methe leakage detection instrumentation.
| |
| m SIn accordanceFrequency with the Surveillance Control Program I!
| |
| I I
| |
| Brunswick Unit 2 3.4-11 Brunsick Uit 2 No. ~~
| |
| .4-11Amendment
| |
| | |
| RCS Specific Activity 3.4.6 ACTIONS (continued' CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours AND B.2.2.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.6.1------------NOTE----------------
| |
| Only required to be performed in MODE 1.
| |
| Verify reactor coolant DOSE EQUIVALENT I-131 specific activity is < 0.2 pCi/gm.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.4-13 Brunsick Uit 2 No. 3
| |
| .4-13Amendment
| |
| | |
| RHR Shutdown Cooling System--Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1-----------NOTE---------
| |
| Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR shutdown cooling isolation pressure.
| |
| Verify one required RHR shutdown cooling subsystem 42-hc'-rs or recirculation pump is operating. l IIn accordance with the Surveillance Frequency Control Program l!
| |
| Brunswick Unit 2 3.4-16 Brunsick Uit 2 .4-16Amendment No.23
| |
| | |
| RHR Shutdown Cooling System--Cold Shutdown 3.4.8 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hour from subsystem in operation. circulating by an alternate discovery of no method. reactor coolant AN D circulation No recirculation pump in AND operation.
| |
| Once per 12 hours thereafter AND B.2 Monitor reactor coolant Once per hour temperature.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling subsystem !2 hc'dr or recirculation pump is operating. "*
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.4-18 Bruswik Uit 3.-18Amendment No. 2334
| |
| | |
| RCS P/T Limits 3.4.9 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C.-----NOTE---------....C.1 Initiate action to restore Immediately Required Action C.2 shall be parameter(s) to within completed if this Condition is limits.
| |
| entered.
| |
| -------- AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for operation. MODE 2 or 3.
| |
| MODES 1, 2, and 3.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.1-----------NOTE- -------
| |
| Only required to be performed during RCS heatup and cooldown operations.
| |
| Verify:
| |
| : a. RCS pressure and RCS temperature are within /
| |
| the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2; and
| |
| : b. RCS heatup and cooldown rates are < 100°F in any 1 hour period.
| |
| ~(continued)
| |
| SSurveillance Frequency SIn accordance with the Control Program Brunswick Unit 2 3.4-20 BrunwickUnit23.-20Amendment No. 22*
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.2-----------NOTE---------
| |
| Only required to be performed during ROS inservice leak and hydrostatic testing.
| |
| Verify: "*n ,-r ,,,i..
| |
| : a. RCS pressure and RCS temperature are within ,
| |
| the applicable'limits specified in Figure 3.4.9-3; 3.4.9-4, or 3.4.9-5, as applicable;
| |
| : b. RCS heatup and cooldown rates are < 30°F in any 1 hour period.
| |
| SR 3.4.9.3 Verify RCS pressure and RCS temperature are within ; nce within the criticality limits specified in Figure 3.4.9-2. j5 minutes prior to o['ntrol rod ithdraa for the purpose of achieving criticality SR 3.4.9.4---------------------NOTE---------------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the bottom head coola tOnce within temperature and the reactor pressure vessel (RPV) 30 minutes prior to coolant temperature is < 145°F. each startup of a recirculation pump (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program 1I Brunswick Unit 2 3.4-21 BrunwickUnit2 3.-21Amendment No. 2,56
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.5-----------NOTE- -------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the reactor coolant Once within temperature in the recirculation loop to be started and 30 minutes prior to the RPV coolant temperature is _<50°F. each startup of a recirculation pump SR 3.4.9.6-----------NOTE- -------
| |
| Only required to be performed when tensioning the reactor vessel head bolting studs.
| |
| Verify reactor vessel flange and head flange ,=-R'-mI'Wte&
| |
| temperatures are >_70°F.
| |
| SR 3.4.9.7-----------NOTE- -------
| |
| Not required to be performed until 30 minutes after ROS temperature _<80°F in MODE 4. I Verify reactor vessel flange and head flange ~0-miR~4tes temperatures are _ 70°F.
| |
| ~6 SR 3.4.9.8-----------NOTE- -------
| |
| Not required to be performed until 12 hours after RC:*
| |
| temperature < 100°F in MODE 4.
| |
| Verify reactor vessel flange and head flange /
| |
| temperatures are _>70°F.*! 1!2-he'-s In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.4-22 Bruswik Uit 3.-22Amendment No. 24?2
| |
| | |
| Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be < 1045 psig.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is _<1045 psig. 42-hc',w'r I
| |
| IIn accordance with the Surveillance Frequency Control Program II II Brunswick Unit 2 3.4-28 BrunwickUnit2 No. 25 I 3.-28Amendment
| |
| | |
| ECCS--Operating 3.5.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME J. Two or more low pressure J.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A or B.
| |
| O__R HPCI System and two or more required ADS valves inoperable.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.5.1 .1 Verifyj, for each ECCS injection/spray subsystem, the 4-days ..
| |
| piping is filled with water from the pump discharge A valve to the injection valve.
| |
| V (continued)
| |
| In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.5-4 BrunwickUni 2 35-4Amendment No. 247
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.2 -------- NOTE----------...............
| |
| Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling isolation pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path,/
| |
| that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.5.1.5--------------------Nticspl e---------------- 4...
| |
| SR3514 Verify eahe RHRSsemrcu pupdicarge osation vas nokd Once-achsaru (contnued BuSwikUi 2... 3.5-5E Amndet o.
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.6 Verify the following ECCS pumps develop the 92-de~s specified flow rate against a system head corresponding to the specified reactor pressure. 4-SYSTEM HEAD CORRESPONDIN(
| |
| NO. OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS >4l00gpm 1 >_113psig LPCI > 14,000 gpm 2 > 20 psig SR 3.5.1.7--------------------NOTE--------- --------
| |
| Not required to be performed until 48 hours a ner reactor steam pressure is adequate to perfor the test.
| |
| Verify, with reactor pressure < 1045 and "945 psig, 24L the HPCI pump unit can develop a flow r te
| |
| >_4250 gpm against a system head cor pnigt reactor pressure.
| |
| SR 3.5.1.8------------NOTE----------------...
| |
| Not required to be performed until 4t hours aft reactor steam pressure is adequate operforrth test. *g Verify, with reactor pressure __1I psghe HPCI 244 menthc pump unit can develop a flow ra .. Ž4 0 gpm against a system head corresponding t_ e or pressure (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.5-6 Brunsick Uit 2 No. a~
| |
| .5-6Amendment
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SU RVEI LLANCE FREQU ENCY SR 3.5.1.9---------------------NOTE---------------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each ECCS injection/spray subsystem actuates 2"!. mRths on an actual or simulated automatic initiation signal.
| |
| SR 3.5.1.10--------------------NOTE----------------
| |
| Valve actuation may be excluded.
| |
| Verify the ADS actuates on an actual or simulated 24-FPcethe automatic initiation signal._.
| |
| SR 3.5.1.11-----------NOTE----------------
| |
| Not required to be performed until 12 hours after reactor steam pressure is adequate to performte test.
| |
| Verify each required ADS valve opens when anu I 4-RnthS injection/spray subsystem is within the I*ir lIn accordance with the Surveillance Frequency Control Program Y/
| |
| Brunswick Unit 2 3.5-7 Brunsick Uit 2 .5-7Amendment No. 2
| |
| | |
| ECCS--Shutdown 3.5.2 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| : 0. Required Action C.2 and 0.1 Initiate action to restore Immediately associated Completion Time secondary containment to not met. OPERABLE status.
| |
| AND D.2 Initiate action to restore one Immediately standby gas treatment subsystem to OPERABLE status.
| |
| AND 0.3 Initiate action to restore Immediately isolation capability in each required secondary containment penetration flow path not isolated.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure coolant injection !2-hc'--'s (LPCI) subsystem, the suppression pool water level is
| |
| >-31 inches. 7 SIn accordance with the I cniud SurvillnceFrequency Control Program Brunswick Unit 2 3.5-9 Amendment No. 22g
| |
| | |
| ECCS-Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE jFREQUENCY Verify, for each required core spray (CS) subsystem, SR 3.5.2.2 the: 1--h\
| |
| : a. Suppression pool water level is >_-31 inches; or b.---------NOTE----------......
| |
| Only one required CS subsystem may take credit for this option during OPDRVs.
| |
| I
| |
| --------------------- I Condensate storage tank water volume is
| |
| _> 228,200 gallons.
| |
| SR 3.5.2.3 Verify, for each required ECCS injection/spray subsystem, the piping is filled with water from the l pump discharge valve to the injection valve.
| |
| SR 3.5.2.4 One LPCI subsystem may be considered OEA during --------
| |
| alignment and operation for decay-*-E-------
| |
| NOTE----..... heatre if capable of being manually realigned and not other~wise inoperable.
| |
| Verify each required ECCS injection/spray subsysln manual, power operated, and automatic valve intl securedflwin thaitisot,is in the correct position.o /A II .~.'
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.5-10 Bruswik Uit 3.-10Amendment No. 2-&
| |
| | |
| ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure. /
| |
| NO. CORRESPONDI NG OF SYSTEM HEAD G TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS >__4100gpm 1 Ž__113 psig LPCI _>9000 gpm 1 _>20 psig SIR 3.5.2.6 -NOTE--------.........----
| |
| Vessel injection/spray may be excluded. I Verify each required ECCS injection/spray subsl actuates on an actual or simulated automatic min signal.
| |
| SR 3.5.2.7 -NOTE----------......
| |
| Instrumentation response time may be assumeq the design instrumentation response time. J Verify the ECCS RESPONSE TIME for each 2"!,months ECCS injection/spray subsystem is within thq SIn accordanceFrequency Surveillance with the Control Program l
| |
| I l
| |
| Brunswick Unit 2 3.5-11 3.-11Amendment BrunwickUnit2 No. 24*
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water from ~-1-d.a...
| |
| the pump discharge valve to the injection valve.
| |
| SR 3.5.3.2 Verify each RCIC System manual, power operated, and automatic valve in the flow path, that is not lock, sealed, or otherwise secured in position, is in the 31di correct position.
| |
| SR 3.5.3.3 ----------- NOTES ...
| |
| : 1. Use of auxiliary steam for the performance o the SR is not allowed./
| |
| : 2. Not required to be performed until 24 hour.,
| |
| reactor steam pressure is adequate to per~
| |
| the test.I Verify, with reactor pressure _>945 psig and <
| |
| psig, the RCIC pump can develop a flow ratej
| |
| >_ 400 gpm against a system head correspon ct reactor pressure.
| |
| I (continued) lIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.5-13 Brunsick Uit 2 .5-13Amendment No. 2
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.4 ------------- NOTES- ------
| |
| : 1. Use of auxiliary steam for the performance of the SR is not allowed with reactor pressure
| |
| _>150 psig.
| |
| : 2. Not required to be performed until 24 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify, with turbine inlet pressure _ 135 psig and < 165 24-meRmhs psig, the RCIC pump can develop a flow rate
| |
| Ž>400 gpm against a system head corresponding to an*
| |
| equivalent reactor pressure.
| |
| 4 SR 3.5.3.5 ------------ NOTE-------.......... -
| |
| Vessel injection may be excluded./
| |
| Verify the RCIC System actuates on an actual or 2"t-menthe simulated automatic initiation signal.
| |
| In accordance with the Surveillance Frequency Control Program I
| |
| h II Brunswick Unit 2 3.5-14 Brunsick Uit 2 .5-14Amendment No. 2
| |
| | |
| Primary Containment 3.6.1.1 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and leakage rate In accordance with testing, except for primary containment air lock testing, the Primary in accordance with the Primary Containment Leakage Containment Rate Testing Program. Leakage Rate Testing Program SR 3.6.1.1.2 Verify drywell to suppression chamber differential 2-eth pressure does not decrease at a rate > 0.25 inch water gauge per minute tested over a 10 minute period at an I initial differential pressure of_> 1.00 psid and
| |
| *<1.25 psid.
| |
| Brunswick Unit 2 3.6-2 Brunsick Uit 2 .6-2Amendment No. 2
| |
| | |
| Primary Containment Air Lock 3.6.1.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1-----------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 3.6.1.1.1.
| |
| Perform required primary containment air lock leakage In accordance with rate testing in accordance with the Primary the Primary Containment Leakage Rate Testing Program. Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary containment air 24m th lock can be opened at a time. I" Brunswick Unit 2 3.6-6 Amendment No. 2-2
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.6.1.3.1 -~~~~NOTES--------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCI Vs that are open under administrative controls.
| |
| Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident 31dJ conditions is closed.
| |
| p.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.6-11 Brunsick Uit 2 .6-11Amendment No. 2
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 1-SR 3.6.1.3.2 ------------ NOTES--------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCI Vs that are open under administrative controls.
| |
| Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing incore probe (TIP) .a1,da...
| |
| shear isolation valve explosive charge.
| |
| SR 3.6.1.3.4 Verify the isolation time of each power operated and In accordance each automatic PCIV, except for MSIVs, is within with the Inservice limits. Testing Program SR 3.6.1.3.5 Verify the isolation time of each MSIV is _>3 seco d In accordance with and
| |
| * 5 seconds. the Inservice Testing Program VI-
| |
| *L*UI ILII lU{E:*LI)
| |
| IIn accordanceFrequency with the Surveillance Control Program I II Brunswick Unit 2 3.6-12 Bruswik Uit 3.-12Amendment No. 24=2
| |
| | |
| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 24-meP4hs position on an actual or simulated isolation signal.
| |
| SR 3.6.1.3.7 Verify a representative sample of reactor 24-mefPths instrumentation line EFCVs actuate to the isolation A
| |
| *position on an actual or simulated instrument line break signal.
| |
| SR 3.6.1.3.8 Remove and test the explosive squib from each sarIn accordance with isolation valve of the TIP System. the Inservice Testing Program SR 3.6.1.3.9 Verify leakage rate through each main steam lirs In accordance with
| |
| _<100 scfh and the combined leakage rate of alfu the Primary main steam lines is < 150 scfh when tested at Containment
| |
| _>25 psig. Leakage Rate Testing Program IIn accordanceFrequency Surveillance Control Program with the i
| |
| Brunswick Unit 2 3.6-13 Brnsic Uit2 .613Amendment No. 2-67
| |
| | |
| Drywell Air Temperature 3.6.1.4 3,6 CONTAINMENT SYSTEMS 3,6.1.4 Drywell Air Temperature LCO 3.6.1.4 Drywell average air temperature shall be < 150°F.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell average air 8 hours temperature not within limit, temperature to within limit.
| |
| B. Required Action and. B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREUENCY SR 3.6.1.4.1 Verify drywell average air temperature is within limit. 2-e~
| |
| I IIn accordance with the Surveillance Frequency Control Program II II Brunswick Unit 2 3.6-14 Brunsick Uit 2 .6-14Amendment No. 2
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.5 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify nitrogen bottle supply pressure of each nitrogen 2A.-hc'-r backup subsystem is __ 1130 psig.
| |
| SR 3.6.1.5.2----------NOTES--------
| |
| : 1. Not required to be met for vacuum breakers tht are open during Surveillances.
| |
| : 2. Not required to be met for vacuum breakers open when performing their intended functi n SR 3.6.1.5.4 Verify the full open setpoint of each vacuu brars2A-,mcthe
| |
| _<0.5 psid.
| |
| SR 3.6.1.5.5 Verify leakage rate of each nitrogen ba kp24-meRt-he SR 3.6.1.5.6 Verify the Nitrogen Backup System s e ir 21-. mcnthe actuation signal.
| |
| Surveillance Frequency Control Program Brunswick Unit 2 3.6-17 Bruswik Uit 3.-17Amendment No. 22
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 NOTE -.
| |
| Not required to be met for vacuum breakers that are open during Surveillances.
| |
| Verify each vacuum breaker is closed. 44-d~e
| |
| ~AND Within 6 hours after any discharge of steam to the
| |
| / suppression chamber from any source AND Within 12 hours following an operation that causes any of the vacuum breakers to open.
| |
| SR 3.6.1.6.2 Perform a functional test of each r breaker. /
| |
| I, AND Within 12 hours after any discharge of IIn accordanceFrequency Surveillance with the steam to the Control Program suppression chamber from the S RVs vacuum{_ .......... ___
| |
| SR 3.6.1.6.3 Verify theis full breaker open
| |
| < 0.5 setpoint of each required psid. 24 ,*Ie~he Brunswick Unit 2 3.6-19 BrunwickUnit2 No. 2-7-9 I 3.-19Amendment
| |
| | |
| Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature is within 2A-heu~s the applicable limits. AN 5 minutes when
| |
| * performing testing that adds heat to the suppression pool l
| |
| SIn accordance with the Surveillance Frequency Control Program f
| |
| l Brunswick Unit 2 3.6-22 Bruswik Uit 3.-22Amendment No. 2-32
| |
| | |
| Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be _>-31 inches and < -27 inches.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression pool 2 hours level not within limits, water level to within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE / FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits. 2Ahc'-rs SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.6-23 Brunsick Uit 2 .6-23Amendment No. 2
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem ~-*4-d.e,,
| |
| manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise\
| |
| SR 3.6.2.3.2 Verify each RHR pump develops a flow rate / 2-deys while operating in the suppression pool cooling m de./
| |
| In accordance with the Surveillance Frequency Control Program I /I Brunswick Unit 2 3.6-25 Brunsick Uit 2 .6-25Amendment No. 2
| |
| | |
| Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LCO 3.6.3.1 The primary containment oxygen concentration shall be < 4.0 volume percent.
| |
| APPLICABILITY: MODE 1 during the time period:
| |
| : a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to
| |
| : b. 24 hours prior to a scheduled reduction of THERMAL POWER to
| |
| < 15% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment oxygen A.1 Restore oxygen 24 hours concentration not within concentration to within limit.
| |
| limit.
| |
| B. Required Action and B.1 Reduce THERMAL 8 hours associated Completion Time POWER to < 15% RTP.
| |
| not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.3.1.1 Verify primary containment oxygen concentration is :7-days within limits. _
| |
| Surveillance SIn accordanceFrequency with the Control Program Brunswick Unit 2 3.6-26 Brunsick Uit 2 .6-26Amendment No. 2
| |
| | |
| Secondary Containment 3.6.4.1 ACTIONS _______
| |
| COMPLETION CONDITION REQUIRED ACTION TIME C. (continued) C.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment equipment hatches 24-meFR-hs are closed and sealed. ,
| |
| SR 3.6.4.1.2 Verify one secondary containment access door is 24-m ^n'lh closed in each access opening.
| |
| SR 3.6.4.1.3 Verify each SGT subsystem can maintain > 0.25 in* 21 months on--
| |
| of vacuum water gauge in the secondary containm STAG GERI*EDI1" for 1 hour at a flow rate <*3000 cfm.
| |
| I-I#
| |
| lIn accordance with the Surveillance Frequency Control Program I
| |
| l II Brunswick Unit 2 3.6-29 Bruswik Uit 3.-29Amendment No. 2-80 i
| |
| | |
| SCIDs 3.6.4.2 ACTIONS (continued) _________________________
| |
| COMPLETION CONDITION REQUIRED ACTION TIME D. Required Action and D.1-----------NOTE--- .
| |
| associated Completion Time LCO 3.0.3 is not applicable.
| |
| of Condition A or B not met - - - -
| |
| during movement of recently irradiated fuel assemblies in Suspend movement of Immediately the secondary containment recently irradiated fuel or during OPDRVs. assemblies in the secondary containment.
| |
| AND D.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.2.1 Verify the isolation time of each automatic SCID is 2A!. menth*
| |
| within limits.
| |
| SR 3.6.4.2.2 Verify each automatic SCID actuates to the isolation 2A-imqethe position on an actual or simulated actuation signal.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.6-32 Bruswik No. 2-8Q I Uit 3.-32Amendment
| |
| | |
| SGT System 3.6.4.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two SGT subsystems E.1 -NOTE---......
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of recently irradiated fuel ..-.......... ..... ...-
| |
| assemblies in the secondary containment or during Suspend movement of Immediately OPDRVs. recently irradiated fuel assemblies in secondary containment.
| |
| AND E.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SOT subsystem for >_10 continuous ~ 1dy hours with heaters operating.I*
| |
| SR 3.6.4.3.2 Perform required SOT filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). i, the VFTP SR 3.6.4.3.3 Verify each SOT subsystem actuates on an actual o simulated initiation signal.
| |
| lIn accordanceFrequency Surveillance with the Control Program III Brunswick Unit 2 3.6-35 Bruswik No. 2-00 I Uit 3.-35Amendment
| |
| | |
| RHRSW System 3.7.1 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual, power operated, and v, .. *-
| |
| automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the I" correct position or can be aligned to the correct position.
| |
| * IIn accordance with the II Surveillance Frequency Control Program y
| |
| II Brunswick Unit 2 3.7-3 BrunwickUni 2 37-3Amendment No. 2-*
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the SW pump suction bay of 24-he'-rs the intake structure is > -6 ft mean sea level.
| |
| SR 3.7.2.2 Verify the water temperature of UHS is < 90.5°F. *4hu SR 3.7.2.3 ----------- NOTE- -------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each SW System manual, power operated, automatic valve in the flow paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| /
| |
| SR 3.7.2.4
| |
| : 1. A single test at the specified Frequency will
| |
| -------- NOTE---------
| |
| satisfy this Surveillance for both units.
| |
| : 2. Isolation of flow to individual components not render SW System inoperable.
| |
| Verify automatic transfer of each DG cooling . Q2~de~Le supply from the normal SW supply to the alter supply on low DG jacket cooling water supply pressure. A (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program I
| |
| I II Brunswick Unit 2 3.7-9 BrunwickUni 2 37-9Amendment No. 240
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.5-----------NOTE-- -------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each required SW System automatic component 2A-imcnths actuates on an actual or simulated initiation signal.
| |
| * I n accordance with the SSurveillance Frequency SControl Program Brunswick Unit 2 3.7-10 BrunwickUnit2 3.-10Amendment No. 2-*
| |
| | |
| CREV System 3.7.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two CREV subsystems--------------NOTE-----------
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of irradiated fuel assemblies- ------------
| |
| in the secondary containment, during CORE E.1 Suspend movement of Immediately ALTERATIONS, or during irradiated fuel assemblies OPDRVs. in the secondary containment.
| |
| OR AND One or more CREV subsystems inoperable due E.2 Suspend CORE Immediately to an inoperable CRE ALTERATIONS.
| |
| boundary during movement of irradiated fuel assemblies AND!
| |
| in the secondary containment, during CORE E.3 Initiate action to suspend Immediately ALTERATIONS, or during OPDRVs.
| |
| OPDRVs.
| |
| SuRvEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREV subsystem for > 15 continuous "gA- ...
| |
| minutes.
| |
| SR 3.7.3.2 Perform required CREV filter testing in accordance naccordance with with the Ventilation Filter Testing Program (VFTP). the VFTP (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.7-13 BrunwickUnit2 No. 2-7-6 I 3.-13Amendment
| |
| | |
| CREV System 3.7.3 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.3 Perform required CRE unfiltered air inleakage testing In accordance with in accordance with the Control Room Envelope the Control Room Habitability Program. Envelope Habitability Program SR 3.7.3.4 Verify each CREV subsystem actuates on an actual or 2"!, mcnths simulated initiation signal. *1 II In accordance with the Surveillance Frequency Control Program
| |
| //
| |
| I Brunswick Unit 2 3.7-14 Bruswik Uit No. 2 I 3.-14Amendment
| |
| | |
| Control Room AC System 3.7.4 ACTIONS (continued) _________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME F. Three control room AC---------NOTE- ----
| |
| subsystems inoperable LCO 3.0.3 is not applicable during movement of irradiated fuel assemblies in the secondary containment, F.1 Suspend movement of Immediately during CORE irradiated fuel assemblies in ALTERATIONS, or during the secondary containment.
| |
| O PD RVs.
| |
| AND F.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND
| |
| !F.3 Initiate actions to suspend Immediately OPDRVs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify each control room AC subsystem has the 24-me*4-hs capability to remove the assumed heat load. ,,
| |
| Surveillance Frequency In accordance with the Control Program tJ Brunswick Unit 2 3.7-17 BrunwickUnit2 3.-17Amendment No. 2-32,
| |
| | |
| Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -NOTE--------------------.......
| |
| Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
| |
| Verify the gross gamma activity rate of the noble gases is < 243,600 IJCi/second after decay of 30 mnts Once within 4 hours afteraŽ_>50%
| |
| increase in the nominal steady 1Surveillance Frequency state fission gas Control Prga release after factoring out increases due to changes in THERMAL POWER level Brunswick Unit 2 3.7-19 Brunsick Uit 2 .7-19Amendment No. 2
| |
| | |
| Main Turbine Bypass System 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete cycle of each main turbine bypass *" '- ...
| |
| valve.
| |
| SR 3.7.6.2 Perform a system functional test. " 4Re*h SR 3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESPONSE'1 24. mc~the TIME is within limits./i IControl Program Brunswick Unit 2 3.7-21 Brunsick Uit 2 .7-21Amendment No. 2
| |
| | |
| Spent Fuel Storage Pool Water Level 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Spent Fuel Storage Pool Water Level LCO 3.7.7 The spent fuel storage pool water level shall be _>19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool A.1-------NOTE- --
| |
| water level not within limit. LCO 3.0.3 is not applicable.
| |
| Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the spent fuel storage pool water level is y
| |
| _>19 feet 11 inches over the top of irradiated fuel ,
| |
| assemblies racks. seated in the spent fuel storage pool =
| |
| l SIn accordance with the Surveillance Frequency Control Program
| |
| /
| |
| II Brunswick Unit 2 3.7-22 Bruswik Uit 3.-22Amendment No. 2438
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7dy availability for each offsite circuit.Ii SR 3.8.1.2------------..............NOTES----------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. A modified DG start involving idling and gradua acceleration to synchronous speed may be us(
| |
| for this SR. When modified start procedurese not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG starts from standby conditions ad ldy achieves steady state voltage > 3750 V and < 4 00 and frequency > 58.8 Hz and < 61.2 Hz.y
| |
| ... *_*(continued)
| |
| In accordance with the Surveillance Frequency Control Program
| |
| ,11 Brunswick Unit 2 3.8-7 Bruswik Uit 3.-7Amendment No. 2-2
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS_(continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.3 ---------- NOTES---------
| |
| : 1. DG loadings may include gradual loading.
| |
| : 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.
| |
| : 5. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each OG is synchronized and loaded and ~A-deys operates for __60 minutes at a load > 2800 kW and
| |
| < 3500 kW.
| |
| SR 3.8.1.4 Verify each engine mounted tank contains Ž 150 gal ofI 31-deys fuel oil. l SR 3.8.1.5 Check for and remove accumulated water from eachA engine mounted tank. / aqye SR 3.8.1.6 Verify the fuel oil transfer system operates to tr aye fuel oil from the day fuel oil storage tank to the mounted tank.
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 2 3.8-8 Bruswik Uit 3.-8Amendment No. 2-*
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.7 ---------- NOTES---------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| 4 *A H*no Verify each DG starts from standby condition and achieves, in < 10 seconds, voltage _>3750 V and frequency >_58.8 Hz, and after steady state conditions are reached, maintains voltage >_3750 V and < 4300 V and frequency > 58.8 Hz and _<61.2 Hz.
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| ! II Brunswick Unit 2 3.8-9 Bruswik No. 2-8 I Uit 3.-9Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUI REMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.8 -~~~~NOTES---------
| |
| : 1. SR 3.8.1.8.a shall not be performed in MODE 1 or 2 for the Unit 2 offsite circuits. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. SR 3.8.1.8.a is not required to be met ifthe unit power supply is from the preferred offsite circuit.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify: 2A -meR÷1h*.
| |
| : a. Automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit; and
| |
| : b. Manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsite circuit.
| |
| (continued)
| |
| Surveillance Frequency lIn accordance with the Control Program Brunswick Unit 2 3.8-10 Brunsick Uit 2 .8-10Amendment No. 2
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.9 -~~~~NOTES---------
| |
| : 1. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 3 and DG 4. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. If performed with the DG synchronized with offsite power, it shall be performed at a power factor < 0.9.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG rejects a load greater than or equal to its associated core spray pump without tripping. 2AmeL h S(continued)
| |
| Surveillance Frequency SIn accordance with the Control Program Brunswick Unit 2 3.8-11 Bruswik Uit 3.-11Amendment No. 2~z*
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 4
| |
| SR 3.8.1.10 -NOTES, A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG's automatic trips are bypassed on an 2A-mePths actual or simulated ECCS initiation signal except:
| |
| : a. Engine overspeed;
| |
| : b. Generator differential overcurrent;
| |
| : c. Low lube oil pressure;
| |
| : d. Reverse power;
| |
| : e. Loss of field; and
| |
| : f. Phase overcurrent (voltage restrained). A I
| |
| (continued)
| |
| IIn accordance with they t Control ProgramI Brunswick Unit 2 3.8-12 Brunsick Uit 2 .8-12Amendment No. 5
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.11 -~~~~NOTES---------
| |
| : 1. Momentary transients outside the load and power factor ranges do not invalidate this test.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG operating at a power factor < 0.9 operates for _>60 minutes loaded to _>3500 kW and
| |
| _<3850 kW. 24 Re. h Si-i.
| |
| SR 3.8.1.12 NOTES. -.
| |
| A single test at the specified Frequency will satisfy thisj Surveillance for both units./
| |
| Verify an actual or simulated ECCS initiation signal I:2A-menthe capable of overriding the test mode feature to returr each OG to ready-to-load operation.
| |
| (continued)
| |
| Surveillance Frequency IIn accordance with the Control Program I
| |
| al Brunswick Unit 2 3.8-13 Bruswik Uit 3.-13Amendment No. 2-*
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.13----------NOTES---------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 3 and DG 4. However, credit maybe taken for unplanned events that satisfy this SR.
| |
| Verify interval between each sequenced load block is 2A-.meinthe within + 10% of design interval for each load sequence relay. ,
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program k/
| |
| II Brunswick Unit 2 3.8-14 Bruswik Uit 3.-14Amendment No. 2-38
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.14 -~~~NOTES---------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 3 and DG 4. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify, on actual or simulated loss of offsite power signal in conjunction with an actual or simulated ECCS initiation signal:
| |
| r2A-R
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses; and
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in
| |
| _<10.5 seconds,
| |
| : 2. energizes auto-connected emergency loads through load sequence relays,
| |
| : 3. maintains steady state voltage >_3750 V and _<4300 V,
| |
| : 4. maintains steady state frequency >_58.8 Hz and
| |
| * 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected emergency loads for
| |
| > 5 minutes.
| |
| In accordance with the Surveillance Frequency Control Program IJ Brunswick Unit 2 3.8-15 BrunwickUnit2 3.-15Amendment No.
| |
| * I
| |
| | |
| Diesel Fuel Oil 3.8.3 SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.8.3.1 For each required OG, verify: " *..
| |
| : a. The associated day fuel oil storage tank ,
| |
| contains _>22,650 gal; and
| |
| : b. The main fuel oil storage tank contains _>20,85q gal per required 0DG.
| |
| SR 3.8.3.2 Verify fuel oil properties of stored fuel oil are testedln In accordance with accordance with, and maintained within the limits , the Diesel Fuel Oil the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.3 Check day fuelforoiland tankremove and theaccumulated main fuel oilwater from storage taps*ach k..
| |
| r SIn accordance with the Surveillance Frequency Control Program m
| |
| I m
| |
| Brunswick Unit 2 3.8-22 BrunwickUnit2 3.-22Amendment No. 22* I
| |
| | |
| DC Sources--Operating 3.8.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Two or more DC electrical power subsystems inoperable.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is _>130 V on float -ey charge. /
| |
| /I SR 3.8.4.2 Verify no visible corrosion at battery terminals and de~s connectors.
| |
| ORR Verify battery connection resistance is _<23.0 pa for inter-cell connections and < 82.8 pohms for ii rack connections.
| |
| SR 3.8.4.3 Verify battery cells, cell plates, and racks show 4g-me~the visual indication of physical damage or abnorm deterioration that degrades performance. ~1 (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program II I Brunswick Unit 2 3.8-24 BrunwickUnit2 3.-24Amendment No. 235
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell !8- ienths and terminal connections are coated with anti-corrosion material. \
| |
| SR 3.8.4.5 Verify each required battery charger supplies 24-~ne~the
| |
| Ž_250 amps at _>135 V for _Ž4 hours.
| |
| SR 3.8.4.6 ------------- NOTES--------
| |
| : 1. The modified performance discharge test in SR 3.8.4.7 may be performed in lieu of the service test in SR 3.8.4.6 once per 60 months.
| |
| : 2. This Surveillance shall not be performed in MODE 1 or 2 for the Unit 2 DC electrical power subsystems. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| I Verify battery capacity is adequate to Supply, and 24-RmeRths maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 2 3.8-25 Brunsick Uit 2 .8-25Amendment No.23
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.7 ------------ NOTES- -------
| |
| : 1. This Surveillance shall not be performed in MODE 1 or 2 for the Unit 2 DC electrical power subsystems. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. A single test at the specified Frequency will.
| |
| satisfy this Surveillance for both units.
| |
| Verify battery capacity is __80% of the manufacturer's rating when subjected to a performance discharge tes*.
| |
| or a modified performance discharge t.est. / 12*mon-ths we battery shows degradation or has reached 85% of the Surveillance Frequec expected life with Control Prga capacity < 100% of manufacturer's rating AND 24 months when battery has reached 85% of the expected life with capacity > 100% of manufacturer's rating Brunswick Unit 2 3.8-26 Brunsick Uit 2 .8-26Amendment No.23
| |
| | |
| Battery Cell Parameters 3.8.6 ACTIONS__________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.
| |
| B. Required Action and B.1 Declare associated battery Immediately associated Completion Time inoperable.
| |
| of Condition A not met.
| |
| O__R One or more batteries with average electrolyte temperature of the representative cells not within limits.
| |
| OR One or more batteries with one or more battery cell parameters not within Category C limits.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 -7 Category A limits.,.,
| |
| (continued)
| |
| SIn accordance with the '
| |
| Surveillance Frequency Control Program Brunswick Unit 2 3.8-31 BruswikUit No. 24=,= I 3.-31Amendment
| |
| | |
| Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY
| |
| *I.
| |
| SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 92" r-,days Category B limits.
| |
| SR 3.8.6.3 Verify average electrolyte temperature of representative cells is Ž_60°F.
| |
| I.
| |
| IIn accordanceFrequency Surveillance with the Control Program I
| |
| I I
| |
| Brunswick Unit 2 3.8-32 BrunwickUnit2 I 3.-32Amendment No. 285
| |
| | |
| Distribution Systems--Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and indicated power 7--de~ys availability to required AC and DC electrical power distribution subsystems.
| |
| SR 3.8.7.2 Verify no combination of more than two power lays conver~sion modules (consisting of either two lightii inverters or one lighting inverter and one plant uninterruptible power supply unit) are aligned to Division IIbus B. I In accordance with the Surveillance Frequency Control Program dI Brunswick Unit 2 3.8-37 BrunwickUnit2 No. 22* I 3.-37Amendment
| |
| | |
| Distribution Systems--Shutdown 3.8.8 ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to suspend Immediately operations with a potential for draining the reactor vessel.
| |
| AND A.2.4 Initiate actions to restore Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and indicated power 7-gays availability to required AC and DC electrical power distribution subsystems.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.8-39 BrunwickUnit2 No. 23 I 3.-39Amendment
| |
| | |
| Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of the following required refueling equipment interlock inputs:
| |
| : a. All-rods-in,
| |
| : b. Refuel platform position,
| |
| : c. Refuel platform fuel grapple, fuel loaded,
| |
| : d. Fuel grapple position,
| |
| : e. Refuel platform frame-mounted hoist, fuel loaded, and
| |
| : f. Refuel platform monorail hoist, fuel loaded.
| |
| "1-In accordance with the l Surveillance Frequency Control Program Brunswick Unit 2 3.9-2 BrunwickUni 2 39-2Amendment No. 22,3,
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
| |
| ACTIONS_________________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-out A,.1 Suspend control rod Immediately interlock inoperable., withdrawal.
| |
| AND A,.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE jFREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in Refuel position. 1 cr
| |
| - I (continued)
| |
| In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.9-3 Brunsick Uit 2 .9-3Amendment No. 2
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.9.2.2------------NOTE--------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn.
| |
| Perform CHANNEL FUNCTIONAL TEST. 7ea-In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.9-4 Brunsick Uit 2 .9-4Amendment No.
| |
| | |
| Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
| |
| APPLICABILITY: When loading fuel assemblies into the core.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control rods A.1 Suspend loading fuel Immediately not fully inserted, assemblies into the core.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SR 3.9.3.1 Verify all control rods are fully inserted ... -r In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.9-5 Brunsick Uit 2 .9-5Amendment No. 2
| |
| | |
| Control Rod OPERABILITY--Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY--Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.
| |
| APPLICABILITY: MODE 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to fully insert Immediately control rods inoperable, inoperable withdrawn control rods.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.5.1---------------------NOTE---------------
| |
| Not required to be performed until 7 days after the control rod is withdrawn.
| |
| Insert each withdrawn control rod at least one notch. -s Verify each withdrawn control rod scram acmltr SR 3.9.5.2 pressure is Ž_940 psig./
| |
| In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 2 3.9-8 Brunsick Uit 2 .9-8Amendment No.23
| |
| | |
| RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 RPV water level shall be _> 23 ft above the top of irradiated fuel assemblies seated within the RPV.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies within the RPV, During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.
| |
| ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not within A.1 Suspend movement of fuel Immediately limit. assemblies and handling of control rods within the RPV.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is > 23 ft above the top of 2 or irradiated fuel assemblies seated within the RPV. , ,
| |
| SIn accordance with the Surveillance Frequency Control Program III j
| |
| II Brunswick Unit 2 3.9-9 BruswikUit 3.-9Amendment No. 24=2
| |
| | |
| RHR--High Water Levei 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is ! cr operating.
| |
| Surveillance lnaccordanceFrequency withth Brunswick Unit 2 3.9-12 Brunsick Uit 2 .9-12Amendment No. 2
| |
| | |
| RHR--Low Water Level 3.9.8 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is hc'-rs operating.
| |
| Surveillance I naccordanceFreq withuenc the Control Program Brunswick Unit 2 3.9-15 Bruswik Uit 3.-15Amendment No. 2-
| |
| | |
| Reactor Mode Switch Interlock Testing 3.10.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Place the reactor mode 1 hour switch in the shutdown position.
| |
| OR A.3.2------NOTE-- --
| |
| Only applicable in MODE 5.
| |
| Place the reactor mode 1 hour switch in the refuel position.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted in core cells !2-he,-rs containing one or more fuel assemblies. /
| |
| Brunswick Unit 2 3.10-5 Amendment No.
| |
| | |
| Single Control Rod Withdrawal--Hot Shutdown 3.10.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.3.2-----------NOTE-........................
| |
| Not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d. 1 requirements.
| |
| Verify all control rods, other than the control rod being !24-.heire SR 3.10.3.3 wihrwaeful all control rods,netdVerify other than the control rod beini *2A. c-he*
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.10-8 Brunsick Uit 2 .10-8Amendment No.23
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One or more of the above B.1 Suspend withdrawal of the Immediately requirements not met with control rod and removal of the affected control rod not associated CRD.
| |
| insertable.
| |
| AND B.2.1 Initiate action to fully insert Immediately all control rods.
| |
| OR B.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.4.2--------------------NOTE---------------
| |
| Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c.1 requirements.
| |
| Verify all control rods, other than the control rod being 2"!.heurs withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| (continued) iIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.10-11 Brunwic Unt 23.1-11Amendment No. 2-*
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.3 Verify all control rods, other than the control rod being 2A-heur-s withdrawn, are fully inserted.
| |
| SR 3.10.4.4-----------NOTE-------------
| |
| Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.b.1 requirements.
| |
| Verify a control rod withdrawal block is inserted.
| |
| I1 IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.10-12 Brunswck Unt No. ~~
| |
| 2 310-12Amendment
| |
| | |
| Single CR0 Removal--Refueling 3.10.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Initiate action to fully insert Immediately all control rods.
| |
| o__R A.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods, other than the control rod 24-heu*s withdrawn for the removal of the associated CR0, are fully inserted. \
| |
| SR 3.10.5.2 Verify all control rods, other than the control rod 2A-.heirs withdrawn for the removal of the associated CR0, in five by five array centered on the control rod withdr f for the removal of the associated CR0, are disarm d SR 3.10.5.3 Verify a control rod withdrawal block is inserted. 2"A.heurs SR 3.10.5.4 Perform SR 3.1.1.1. According to SR 3.1.1.1 m
| |
| LIn Surveillance Frequency accordance with the Control Program l
| |
| m (continued)
| |
| Brunswick Unit 2 3.10-14 Brunwic Unt 23.1-14Amendment No. 2-&3
| |
| | |
| Single CR0 Removal--Refueling 3.10.5 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.5 Verify no other CORE ALTERATIONS are in progress. 124-hef
| |
| -L h
| |
| IIn Surveillance Frequency accordance with the Control Program tj' d
| |
| Brunswick Unit 2 3.10-15 Brunwic Unt 23.1-15Amendment No. 23,2
| |
| | |
| Multiple Control Rod Withdrawal--Refueling 3.10.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Initiate action to fully insert Immediately all control rods in core cells containing one or more fuel assemblies.
| |
| OR A.3.2 Initiate action to satisfythe Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.6.1 Verify the four fuel assemblies are removed from core 24-heurs cells associated with each control rod or CRDl removed.
| |
| SR 3.10.6.2 I 24-heers Verify all other control rods in core cells containing or more fuel assemblies are fully inserted.
| |
| SR 3.10.6.3-----------NOTE---------------..
| |
| Only required to be met during fuel loading.
| |
| Verify fuel assemblies being loaded are in compliantc with an approved spiral reload sequence. p 24-h* ,r IIn accordanceFrequency Surveillance with the Control Program I
| |
| IM Brunswick Unit 2 3.10-17 BrunwickUnit2 3.0-17Amendment No. 223*
| |
| | |
| SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SU RVEI LLANCE FREQU ENCY SR 3.10.8.2--------------------NOTE---------------
| |
| Not required to be met if SR 3.10.8.3 satisfied.
| |
| Perform the MODE 2 applicable SRs for LCO 3.3.2.1, According to the Function 2 of Table 3.3.2.1-1. applicable SRs SR 3.10.8.3--------------------NOTE---------------
| |
| Not required to be met if SR 3.10.8.2 satisfied.
| |
| Verify movement of control rods is in compliance with During control rod the approved control rod sequence for the SDM test by movement a second licensed operator or other qualified member of the technical staff.
| |
| SR 3.10.8.4 Verify no other CORE ALTERATIONS are in progress. 12 hc'-rs (continued)
| |
| Control Program Brunswick Unit 2 3.10-22 Brunswck Unt2 310-22Amendment No. 2~
| |
| | |
| SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.5 Verify each withdrawn control rod does not go to the Each time the withdrawn overtravel position. control rod is withdrawn to "full out" position AND Prior to satisfying LCO 3.10.8.c requirement after work on control rod or CRD System that could affect coupling SR 3.10.8.6 Verify CRD charging water header pressure _> 940 ds psig.
| |
| Surveillance Frequency In accordance wt h Control Program tJ Brunswick Unit 2 3.10-23 BrunwickUnit2 3.0-23Amendment No. 233
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals Control Room Envelope Habitability Procqram (continued)
| |
| : e. The quantitative limits on unfiltered air inleakage into the ORE. 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 in leakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of ORE 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 ORE habitability, determining ORE unfiltered inleakage, and measuring ORE pressure and assessing the ORE boundary as required by paragraphs c and d, respectively.
| |
| l 5.5.14 Surveillance Frequency Control Progiram This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| : 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.
| |
| Brunswick Unit 2 5.0-17a Brunwic Unt No. 2-76 I 25.017aAmendment
| |
| | |
| BSEP 15-0101 Enclosure 5 Marked-up Technical Specification Bases Pages - Unit 1 (For Information Only)
| |
| | |
| Brunswick Unit 1 Bases inserts for TSTF-425, Revision 3 Insert 1 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| Insert 2 The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
| |
| | |
| Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 (continued)
| |
| REQUIREMENTS new control rod or a control rod from another core location. The 24 hour interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and predicted core keff can be made. For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at >_75% RTP have been obtained.
| |
| Also, core reactivity changes during the operating cycle. Thorofor~eh to '-ariations in core roactivity. The comparison of monitored and predicted core reactivity requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results. Therefore, the comparison at this Froquency is only done when in MODE 1. The cor..weight,+ ton,,, (T) in MWDnn, reflects REFERENCES 1. UFSAR, Section 3.1.2.3.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 ..-
| |
| B3.1.2-5 eiinN.3 Revision No. 31 I
| |
| | |
| Control Rod OPERABILITY B 3.1.3 BASES ACTIONS E.1 (continued)
| |
| Ifany Required Action and associated Completion Time of Condition A, C, or D are not met, or there are nine or more inoperable control rods, 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 12 hours.
| |
| This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 8.75% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns.
| |
| Control rod position may be determined by the use of OPERABLE reed switch position indicators (including "full-in" or "full-out" indication), by moving control rods to a position with an OPERABLE reed switch indicator, or by the use of other appropriate methods. Th-T424-hei*
| |
| Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. As noted, SR 3.1.3.2 is not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM. This Note acknowledges that the (continued)
| |
| Brunswick Unit 1 B3.1.3-7 Brunwic Uni No. 63 I 1 3.13-7Revision
| |
| | |
| Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.2 (continued)
| |
| REQUIREMENTS control rod must first be withdrawn and THERMAL POWER must be increased to above the LPSP of the RWM before performance of the Surveillance. Thus the Note avoids potential conflicts with SR 3.0.3 and SR 3.0.4. The Surveillance is not required to be performed when I THERMAL POWER is less than or equal to the LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the BPWS (LCO 3.1.6) and the RWM (LCO 3.3.2.1). While performance of the SR is exempted during this condition, the SR must still be met.. T-he I c~hngcs inq CRD pcrf, rmanco. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) *ust be made and appropriate action taken. Insrt 1I SR 3.1.3.3 I Verifying that the scram time for each control rod to notch position 06 is
| |
| < 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (S DV) Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
| |
| (continued)
| |
| I I Brunswick Unit 1 B 3.1.3-8 Brunwic Uni No. 63 I I 3.13-8Revision
| |
| | |
| Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (co ontinued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." With more than 7.5% of the sample declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all surveillances) exceeds the LCO limit.
| |
| For planned testing, the control rods selected for the sample should be different for each test. This test is performed for each control rod in the sample from its fully withdrawn position. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample4.]
| |
| *. Tho111 200LIV day. FrcIu.n4 i sIlIV**U b IoI on oprItI ngIUIL~
| |
| clpclIsns VVth.l t Accumula'tors."
| |
| SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate the affected control rod is still within acceptable limits.
| |
| This test is performed for each affected control rod from its fully withdrawn position. In lieu of actually initiating a scram for each affected control rod, testing that adequately demonstrates the scram times are within acceptable limits is allowed to satisfy this SR. The test may include any series of sequential, overlapping, or total steps so the entire scram (continued)
| |
| Brunswick Unit 1 B 3.1.4-5 Brunwic Uni No. 42 I 1 3.14-5Revision
| |
| | |
| Control Rod Scram Accumulators B 3.1.5 BASES (continued)
| |
| SURVEILLANCE REQUIREMENTS SR 3.1.5.1 / [periodically SR 3.1.5.,1,rquires that the control rod scram accumulator pressure be checked **7dysto ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of approximately 1100 psig.
| |
| Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that appropriate action is taken if significant degradation in scram capability occurs. This Surveillance may be performed by verification of absence of the common scram accumulator low.presure. alar...
| |
| REFERENCES 1. UFSAR, Section 4.2.1.1.8.
| |
| : 2. UFSAR, Section 4.3.2.
| |
| : 3. UFSAR, Chapter 15.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B3.1.5-5 Brunwic Uni No. 31 I 1 3.15-5Revision
| |
| | |
| Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued)
| |
| When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor must be manually scrammed within 1 hour. This ensures the reactor is shut down and, as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.
| |
| SURVEILLANCE SR 3.1.6.1 tproial REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at-a 2*)1,hour
| |
| ... rc,.......n-c, to ensure the assumptions of the CRDA analyses are me÷ Thc, 21 hour F'rcqucn , wac dcv--clopcd co-;ic;in that thc prim,'
| |
| IInsert I (LC 3.3..1).. , ,,hich, provides control rod blocks to enforce the required sequence and is re *red to be OPERABLE when operating at
| |
| < 8.75% RTP. /
| |
| IThe RWMI REFERENCES 1. UFSAR, Section 15.4.
| |
| : 2. XM-NF-80-19(P)(A), Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis," (as identified in the COLR).
| |
| : 3. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO/MICROBURN-B2," (as identified in the COLR).
| |
| : 4. NEDE-2401 1-P-A-i11-US, General Electric Standard Application for Reactor Fuel, Supplement for United States, Section 2.2.3.1, November 1995.
| |
| : 5. NRC Safety Evaluation Report, Acceptance for Referencing of Licensing Topical Report NEDE-2401 1-P-A, General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17; December 27, 1987.
| |
| : 6. UFSAR, Section 4.3.2.5. II
| |
| : 7. NUREG-0800, Section 15.4.9, Revision 2, July 1981.
| |
| : 8. NEDO-21778-A, Transient Pressure Rises Affected Fracture
| |
| ]
| |
| Toughness Requirements for Boiling Water Reactors, December 1978.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.1.6-4 Brunwic Unt No. 58 I I 3.6-4Revision
| |
| | |
| SLC System B 3.1.7 BASES ACTIONS C.1 (continued)
| |
| If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours.
| |
| The allowed Completion Time of 12 hours 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 3.1.7.1, SR 3.1.7.2. and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 arc 21 hou-r Surv*eillances verifyiaeg certain characteristics of the SLC System (e.g., the volume and temperature of the borated solution in the storage tank), thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure that the proper borated solution volume and temperature, including the temperature of the pump suction and.
| |
| discharge piping up to the SLC injection valves, are maintained.
| |
| Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank or in the pump suction piping. The temperature versus concentration curve of Figure 3.1.7-2 ensures that a
| |
| [ Inset 2
| |
| * 5°F margin will be maintained above the saturation temperature. T-he SR 3.1.7.4 SR 3.1.*7.4 verifies the continuity of the explosive charges in the SLC injection valves to ensure that proper operation will occur if required.
| |
| Other administrative controls, such as those that limit the shelf life of the
| |
| *i i, ** *, iii l,*y In ert I I (continued)
| |
| Brunswick Unit 1 B 3.1.7-4 Brunwic Uni No. 31 I 1 3.17-4Revision
| |
| | |
| SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.5 REQUI REMENTS (continued) This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron (measured in weight % sodium pentaborate) exists in the storage tank. SR 3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. SR 3.1.7.5 must also be performed anytime the temperature is restored to within the limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred during the time period temperature was outside the limits of the Figure.
| |
| rclati-'cly slow -'ariation of boron conccntration bctween Sur-,cillanccs.
| |
| [Insrt 1SR 3.1.7.6 Demonstrating that each SLC System pump develops a flow rate
| |
| > 41.2 gpm at a discharge pressure > 1190 psig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance.
| |
| Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is in accordance with the Inservice Testing Program.
| |
| SR 3.1.7.7 This Surveillance ensures that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch
| |
| (.continued_).
| |
| Brunswick Unit 1 B 3.1.7-5 Brunwic Uni No. 31 I 1 3.17-5Revision
| |
| | |
| SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 (continued)
| |
| REQUIREMENTS that has been certified by having one of that batch successfully fired. T-he pump ana cxpio~ivc vaivo icstca ~nouia oc aitornaico ~ucri inai ooin copl 4to flow
| |
| ... th. arc.t....o; 8mnh a ontig2 ot intcr'-as. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPVzA SR 3.1.7.8 IIsr Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water. Isotopic tests on the granular sodium pentaborate to verify the actual B-I10 enrichment must be performed prior to addition to the SLC tank in order to ensure that the proper B-10 atom percentage is being used.
| |
| REFERENCES 1. 10 CFR 50.62.
| |
| : 2. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, Final Report, February 1, 1995.
| |
| : 3. UFSAR, Section 9.3.4.
| |
| : 4. 10 CFR 50.67
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 ..-
| |
| B 3.1.7-6 eiinN.3 No. 34 Revision
| |
| | |
| SDV Vent and Drain Valves B 3.1.8 BASES (continued)
| |
| SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2, at which time the valves may be closed intermittently under administrative control) to allow for drainage of the SDV piping. Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended functions during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position.
| |
| L[Insrt1 Th 31L__J dayL Frcguo.... is bascd on enginccring judgmcnt and is
| |
| ................. inFpoi~cdural controls govcrning valvc opcration, which SR 3.1.8.2 During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping. Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram../ Th.... 31 day , .....nc is SR 3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The closure time of 30 seconds after receipt of a scram signal is based on the bounding leakage case evaluated in the accident analysis (Ref. 2).
| |
| Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3, "Control Rod - OPERABILITY," overlap this
| |
| [iIset 1-.. Surveillance to provide complete testing of the assumed safety function.
| |
| SThc-- 21 mo'nth Frc........ is.ba.,scdf an* thc n... to'. p,- rform,,-, thi.
| |
| (continued)
| |
| Brunswick Unit 1 B3.1.8-4 Brunwic Uni No. 31 I 1 3.18-4Revision
| |
| | |
| SDV Vent and Drain Valves B 3.1.8 BASES SUVILL\/-IANCEI~- SR 3."1.3 (c"/".nt~inued),A S.....l nc u..d. r t4. hc c..ndition.s. th..t apply:" dur;ing a* plant outage. and thc potential.f.. .an un_ anned tran.i.n if'*t*he Su.....;ln... wcopror me th.th...rea..tor.at..p..er.
| |
| ... eaigeprec hapeontaedfthese component will!! usually pass th*e Sureilanc when... " performed at the*
| |
| REFERENCES 1. 10 CFR 50.67.
| |
| : 2. NUREG-0803, Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping, August 1981.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.1.8-5 Brunwic Uni No. 31 I I 3.18-5Revision
| |
| | |
| APLHGR B 3.2.1 BASES (continued)
| |
| APPLICABILITY The APLHGR limits are derived from LOCA analyses that are assumed to occur at high power levels. Studies and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. At THERMAL POWER levels < 23% RTP, the reactor is operating with substantial margin to the APLHGR limits. For consistency with the 2.1.1.1 SL, this power level was selected for LCO applicability.
| |
| ACTIONS A.__
| |
| If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met.
| |
| Therefore, prompt action should be taken and continued to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 4 hour Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.
| |
| If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE REQUIREMENTS SR 3.2.1.1 [periodicall.y k APLHGRs are required to be initially calculated within,12 *urs after THERMAL POWER is Ž__23% RTP and then eciery-2A, hours thereafter.
| |
| They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. TFhe norm.l pcaion... ; The 12 hour allowance after THERMAL POWER
| |
| > 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.
| |
| (continued)
| |
| Brunswick Unit 1 B3.2.1-2 Brunwic Uni No. 77 I I 3.21-2Revision
| |
| | |
| MCPR B 3.2.2 BASES ACTIONS A.1_
| |
| If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.
| |
| Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 4 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.
| |
| B.1__
| |
| If the MCPR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE REQUIREMENTS SR 3.2.2.1 iperiodically The MCPR is required to be initially calculated within 12 h rs after THERMAL POWER is Ž>23% RTP and then eve -heHs"thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis..:Fhe-24-heuw Becus theotraoansienanalysi TakEsMA crWEdi for conservTPism incheed sca speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.
| |
| SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is then determined based either on the applicable limit associated with scram times of LCO 3.1.4, "Control Rod Scram Times," or the realistic (nominal) scram times. The scram time dependent MCPR limits are contained in the COLR. This determination must be performed within 72 hours after each set of control rod scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle. The 72 hour Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.
| |
| (continued) I B 3.2.2-3 Revision No. 58 Brunswick Unit 1I Brunswick B 3.2.2-3 Revision No. 58 I
| |
| | |
| LHGR B 3.2.3 BASES ACTIONS A.__
| |
| If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis in not met. Therefore, prompt action should be taken to restore the LHGR(s) to within its required limits such that the plant is operating within analyzed conditions. The 4 hour Completion Time is normally sufficient to restore the LHGR(s) to within its limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification.
| |
| B.._
| |
| If the LHGR cannot be restored within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER is reduced to < 23% RTP within 4 hours.
| |
| The allowed Completion Time is reasonable, based on operating experience, to reduce THERAL POWER to < 23% RTP in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SIR 3.2.3.1 periodically k REQUIREMENTS The LHGRs are required to be initially calc ated within 12 hours after THERMAL POWER is > 23% RTP and then e el.-'2A,-heurc thereafter.
| |
| They are compared with the LHGR limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis1 T[he 2...hour. F"quc i...ba...od on both... ong.n...n judgm. ent nd
| |
| ..... nsr normal* .p.r.tion. The 12 hour allowance after THERMAL POWER
| |
| --23% RTP is acceptable given the large inherent margin to operating limits at low power levels.
| |
| REFERENCES 1. UFSAR, Chapter 4.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. NUREG-0800, Section 4.2.11 A.2(g), Revision 2, July 1981.
| |
| : 4. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).
| |
| : 5. 10 CFR 50.36(c)(2)(ii)
| |
| Brunswick Unit 1 B 3.2.3-3 Brunwic Uni No. 58 I I 3.23-3Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE That analysis demonstrated that the 6 hour testing allowance does not REQUIREMENTS significantly reduce the probability that the RPS will trip when necessary.
| |
| (continued)
| |
| SR 3.3.1.1.1 (Not used.)
| |
| SR 3.3.1.1.2 Performance of the CHANNEL CHECK ....... cc 24,; .... enurs.ha a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or 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 CHANNEL CHECK for APRM functions includes a step to confirm that the automatic self-test functions for the APRM and RBM chassis are still operating.
| |
| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| [insert channcl fail'-rc ia rac. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCD.
| |
| (continued)
| |
| Brunswick Unit 1 ...- 0ReiinN.3 B 3.3.1.1-30 Revision No. 31 I
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 REQUIREMENTS --[Insert 1I (continued) To ensure that the APRMs are a curately indicating the true core average power, the APRMs are adjuste */o conform to the reactor power calculated from a heat balance. ,,hrrc Frcqucncy of oncc pcr 7 days i A restriction to satisfying this SR when < 23% RTP is provided that requires the SR to be met only at > 23% IRTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 23% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At > 23% RTP, the Surveillance is required to have been satisfactorily performed 'w'thiln-the
| |
| ,,a.t-I-day7,,*x,
| |
| *.. in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 23% if the 7-day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 23% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
| |
| SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7-day Frequency is not met per SR 3.0.2. In this event, the SR must be
| |
| [Insert 1I'- performed within 12 hours after entering MODE 2 from MODEl1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SIR.
| |
| A Frcqunc of.7*d.,.,z p.,vidcc.an vv. tbl
| |
| .... .. c f ytm vr (continued*
| |
| Brunswick Unit 1 B 3.3.1.1-31 BrunwickUnit1 B.3.11-31Revision No. 31
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.5 REQUIREMENTS (continued) There are four pairs of RPS automatic scram contactors with each pair associated with an RPS scram test switch. Each pair of scram contactors is associated with an automatic scram logic channel (Al, A2, B1, and B2).
| |
| The automatic scram contactors can be functionally tested without the necessity of using a scram function trip. Surveillance Frequency extensions for RPS Functions, described in Reference 11, are allowed provided the automatic scram contactors are functionally tested weekly.
| |
| This functional test is accomplished by placing the associated RPS scram test switch in the trip position, which will deenergize a pair of RPS automatic scram contactors thereby tripping the associated RPS logic channel. The RPS scram test switches were not specifically credited in the accident analysis. However, because the Manual Scram Functions at BNP were not configured the same as the generic model in Reference 11, the RPS scram test switches were evaluated in Reference 12.
| |
| Reference 12 concluded that the Frequency extensions for RPS Functions are not affected by the difference in RPS configuration since
| |
| [periodically each automatic RPS channel has a test switch which is functionally the
| |
| '\same as the manual scram switches in the generic model. As such,
| |
| *1a functional test of each automatic scram contactor is required to be performed-eo., d.,,° ..-
| |
| SR 3.3.1.1.6 and SR 3.3.1.1.7 Isr1I These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.
| |
| The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned fr~om the SRMs to the IRMs.
| |
| The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that (continued)
| |
| Brunswick Unit 1 B 3.3.1.1-32 Brunwic Unt No. 31 I I 3..1.-32Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued)
| |
| REQUIREMENTS the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRM readings have doubled before the SRMs have reached the high-high upscale trip.
| |
| As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2).
| |
| If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable.
| |
| IInsert I A Frcqucncy of 7 dayc ic rca~onablc I3aee~-eR- cnginczring judgmcnt and SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM Insert 1 s, ,c'-d bc ad*i'~c in÷..
| |
| ivcmc
| |
| ....... . ti"n to incrcasas in" Iiccnscd ratcd corc t cr...p.......th orc--vr
| |
| . ag... d.... t..............cc..... tod
| |
| -4
| |
| ~i~* ~.i ~,--,- -* *** t*~
| |
| thc LPRM dctcctor uncortainty rcmain~ Icac than that u~cd to dctcrminc powcr distribution unccrtainty applicd in thc calculation of thc Safct'~,' Limit n~. i,:lu~xrI In ~~i~ruIuti SR 3.3.1.1.9 and SR 3.3.1.1.12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. Th,,,,o* 924"da
| |
| * of,.
| |
| I Insert 2 *.*
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.1.1-33 Brunwic Unt No. 71 I I 3..1.-33Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 (continued)
| |
| REQU IREMENTS Thc, 24/ monhlFlquncy o f SR 3.3.1 .1.12 k* ba~cd on thc nccd to SR 3.3.1.1.10 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable ifthe trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate l Iser 1]*,,,,*\setpoint methodology.
| |
| Thc^ Frcqucncy of2days! bas.cd on tho rcl-'bil:.t-, ,anal,,is o Rcf....... 11.
| |
| SR 3.3.1.1.11 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The scope of the APRM CHANNEL FUNCTIONAL TEST is that which is necessary to test the hardware. Software controlled functions are tested only incidentally. Automatic self-test functions check the EPROMs in which the software-controlled logic is defined. Changes in the EPROMs will be detected by the self-test function and alarmed via the APRM trouble alarm. SR 3.3.1.1.2 for the APRM functions includes a step to confirm that the automatic self-test function is still operating.
| |
| (continued)
| |
| Brunswick Unit 1 ...-
| |
| B 3.3.1.1-34 4ReiinN.3 Revision No. 31 ]
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 (continued)
| |
| REQUIREMENTS The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b and the auto-enable portion of Function 2.f only), the 2-Out-Of-4 Voter channels, and the interface connections into the RPS trip systems from the voter channels.
| |
| Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. *Thc 14 day Fr,,cq....... of afd-6 Insert 11 (NOTE: The actual voting logic of the 2-Out-Of-4 Voter Function is tested as part of SR 3.3.1.1.15. The auto-enable setpoints for the OPRM Upscale trip are confirmed by SR 3.3.1.1.19.)
| |
| A Note is provided for Function 2.a that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
| |
| A second Note is provided for Functions 2.b and 2.f that clarifies that the CHANNEL FUNCTIONAL TEST for Functions 2.b and 2.f includes testing of the recirculation flow processing electronics, excluding the flow transmitters.
| |
| SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.
| |
| CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The CHANNEL CALIBRATION for Functions 5 and 8 should consist of a physical inspection and actuation of the associated position switches.
| |
| Note 1 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7-day (continued)
| |
| Brunswick Unit 1 B 3.3.1.1-35 Brunwic Unt I 3..1.-35Revision No. 31 J
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.13 (continued)
| |
| REQUI REMENTS calorimetric calibration (SR 3.3.1.1.3) and the 2000-EF-P4 LPRM calibration against the TIPs (SR 3.3.1.1.8).
| |
| A second Note is provided that requires the IRM SIRs to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This Note allows entry into MODE 2 from MODE 1 ifthe associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SIR.
| |
| A third note is provided that requires that the recirculation flow (drive flow) transmitters, which supply the flow signal to the APIRMs, be included in the SR for Functions 2.b and 2.f. The APRM Simulated Thermal Power--
| |
| High Function (Function 2.b) and the OPRM Upscale Function (Function 2.f) both require a valid drive flow signal. The APRM Simulated Thermal Power--High Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS. A CHANNEL CALIBRATION of the APRM drive flow signal requires both calibrating the drive flow transmitters and the processing hardware in the APRM equipment. SR 3.3.1.1.18 establishes a valid drive flow/core flow relationship. Changes throughout the cycle in the drive flow/core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power--High Function and the OPIRM Upscale Function.
| |
| [Insert 1 21 month calibration intcr--l in thc dctcrmination of thc m...gn....de of SR 3.3.1.1.14 (Not used.)
| |
| (continued' Brunswick Unit 1 B 3.3.1.1-36 Brun wic Unt I 3. No. 71 I
| |
| -36Revision
| |
| .1.
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic and simulated automatic operation for a specific channel. The functional testing of control rods (LCO 3.1.3),
| |
| and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-Out-Of-4 Voter channel inputs to check all combinations of two tripped inputs to the 2-Out-Of-4 logic in the voter channels and APRM related redundant RPS LIserl.-* relays.
| |
| Trhe- 24 mnnth Fronu-cnnv !shrinrd 1n the nnr~1 tn ncnrirm this Sur....illan unvrh
| |
| ... conditions....th.t...pply duing a.pl.t.ouag...d..h SR 3.3.1.1.16 This SR ensures that scrams initiated from the Turbine Stop Valve--
| |
| Closure and Turbine Control Valve Fast Closure, Control Oil Pressure--
| |
| Low Functions will not be inadvertently bypassed when THERMAL POWER is > 26% RTP. This is satisfied by calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the Allowable Value and the actual setpoint.
| |
| Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during an in-service calibration at THERMAL POWER _> 26% RTP to ensure that the calibration is valid.
| |
| If any bypass channel setpoint is nonconservative (i.e., the Functions are bypassed at _>26% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve--Closure and Turbine Control Valve Fast Closure, Control Oil Pressure--Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (non-bypass). If placed in the (continued' Brunswick Unit 1 B 3.3.1.1-37 Brunwic Unt No. 31 I I 3..1.-37Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued)
| |
| REQUIREMENTS non-bypass condition, this SR is met and the channel is considered OPERABLE.
| |
| I[l*t31t I *T6* i i i,,,., iC.*.,,*.*.
| |
| .*li*f,..'l*.,
| |
| i *.,.n,*,.t*,,..iivj
| |
| *4: CL,.,*
| |
| of,,24*°m,,,nths*is based on ong',nocring ju,.,m..... and iUI;.;U;;;;.¥ Ui LiiU ;*UIiiUUiiU;;L*.
| |
| SR 3.3.1.1.17 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 13.
| |
| RPS RESPONSE TIME for the APRM 2-Out-Of-4 Voter Function (2.e) includes the output relays of the voter and the associated RPS relays and contactors. (The digital portion of the APRM and 2-Out-Of-4 Voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration checks the time base of the digital electronics.
| |
| Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time.)
| |
| Note 2 states the response time of the sensors for Functions 3 and 4 may be assumed in the RPS RESPONSE TIME test to be the design sensor response time. This is allowed since the sensor response time is a small part of the overall RPS RESPONSE TIME (Ref. 14).
| |
| LInert
| |
| ..hann.... sp........in.T.blc 3.3.1.1 1 for"thc,,M.,IV' Clsr-s,..,, Function'-,,,,*,,,÷
| |
| . This,
| |
| {-rcgusnoy
| |
| ,,-..dto.... is Dasce on mc' pr...u...an RPSlo'..... rrc...a'ir ~nsnins scram signal.
| |
| OT tnc variau~ ~nn~i~
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.1.1-38 BrunwickUnit1 B.3.11-38Revision No. 31
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVIDrLL IANCPE ~K &1.i.L1J (continued)
| |
| REQUIIREM ENI/
| |
| f"KTS Notc 4 allows the STAGGERED TEST BASIS Frcgucney for Function 2.c to be doterminod based on 8 channels rather than thc 4 actual 2 Out Of 4
| |
| ~In+a.-akan.,nIe. TkaAu.An...4a.4,-~,4,.4.arn+kaOflu.+flA~Infa,~
| |
| V 'it'i '.1 *CAI II .Ili. I i iiJ l.4lII.All.4S1 .It'LtjJULttJ il.Pil II*fl
| |
| ., ra nr.nr' ,,4 ararLnnrLnt LL ....... \-- L,. ..........
| |
| scprat .. ha...nc..... for applicatio of SR,3.3*.1,.1.17, so n - . Trhc, notc v
| |
| ~~Voter Channel
| |
| * OPRM APIIFa OPRM APRM OPRM, PARM OPRM APRM 3 rd X
| |
| /- ~x xx out.outs,, fromte... 2 Out.. Of 4 Voe. cane.ortatFn.in and each-' at, I -I ..........
| |
| Tfl'' ~ Li Li'. fl flflflfll antI'' fl1 nI *I~ a Li Li'.
| |
| \L,;i.iL L'.
| |
| ,i L, ;L,, ';.- -- ' i*J itI.
| |
| *O L'.
| |
| ,,3 ,L. ,A L-'i 4*,'.,5(4
| |
| ~
| |
| ranIane-.n
| |
| 'i91.ILi*
| |
| :1 ,rf~aA 1(43 JLILit*I*'iL.I Ii) k., ~
| |
| I "ii'.' Li* *'i'iLi 4ltd C ......A 5.45 n.J 40 I (continued)
| |
| Brunswick Unit 1 B 3.3.1.1-39 Brunwic Unt No. 31 I I 3..1.-39Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (conti....c,)
| |
| REQU IREMENTS Th'c 2)4 month Frcg,,n, ;* c -"n.i-tont with thc BNP*I rfucling ccic and* i" b...d upon plan ..... rat..... u---cr......., .. hi..h..ho ...tha adm fiuo of intrmotio cmcnnt cusngsciusro pn tim SR 3.3.1.1.18 The APRM Simulated Thermal Power--High Function (Function 2.b) uses drive flow to vary the trip setpoint. The OPRM Upscale Function (Function 2.f) uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS. Both of these Functions use drive flow as a representation of reactor core flow. SR 3.3.1.1.13 assures that the drive flow transmitters and processing electronics are calibrated. This SR adjusts the recirculation drive flow scaling factors in each APRM channel to provide the appropriate drive flow/core flow alignment.
| |
| The Frequency of once within 7 days after reaching equilibrium conditions following a refueling outage is based on the expectation that any change in the core flow to drive flow functional relationship during power operation would be gradual and the maintenance on the Recirculation System and core components which may impact the relationship is expected to be performed during refueling outages. The 7 day time period after reaching equilibrium conditions is based on plant conditions required to perform the test, engineering judgment of the time required to collect and analyze the necessary flow data, and engineering judgment of the time required to enter and check the applicable scaling factors in each of the APRM channels. The 7-day time period after reaching equilibrium conditions is acceptable based on the relatively small alignment errors expected, and the margins already included in the APRM Simulated Thermal Power--
| |
| High and OPRM Upscale Function trip-enable setpoints.
| |
| SR 3.3.1.1.19 This surveillance involves confirming the OPRM Upscale trip auto-enable setpoints. The auto-enable setpoint values are considered to be nominal values as discussed in Reference 21. This surveillance ensures that the OPRM Upscale trip is enabled (not bypassed) for the correct values of
| |
| __ (continued*
| |
| Brunswick Unit 1 B 3.3.1.1-40 Brunwic Unt No. 31 I 1 3..1.-40Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.19 (continued)
| |
| REQUIREMENTS APRM Simulated Thermal Power and recirculation drive flow. Other surveillances ensure that the APRM Simulated Thermal Power and recirculation drive flow properly correlate with THERMAL POWER (SIR 3.3.1.1.3) and core flow (SR 3.3.1.1.18), respectively.
| |
| In any auto-enable setpoint is nonconservative (i.e, the OPRM Upscale trip is bypassed when APRM Simulated Thermal Power _> 25% and recirculation drive flow _< 60%), then the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted to place the channel in a conservative condition (not bypassed). If the OPRM Upscale trip is placed in the not-bypassed condition, this SR is met and the channel is considered OPERABLE.
| |
| Thc Frogucncy of 21 months is bascd on on~inccrinqw jud~mont
| |
| . v and rcliabilit,, of#4.t-, comonc'nts.L REFERENCES 1. UFSAR, Section 7.2.
| |
| : 2. UFSAR, Chapter 15.0.
| |
| : 3. UFSAIR, Section 7.2.2.
| |
| : 4. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. NEDO-23842, Continuous Control Rod Withdrawal in the Startup Range, April 18, 1978.
| |
| : 7. UFSAR, Section 5.2.2.
| |
| : 8. UFSAR, Appendix 5A.
| |
| : 9. UFSAR, Section 6.3.1.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.1.1-41 Brunwic Unt No. 36 I I 3..1.-41Revision
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES REFERENCES 22. General Electric Nuclear Energy Letter NSA 01-212, (continued) DRF C51-00251-00, A. Chung (GE) to S. Chakraborty (GE),
| |
| "Minimum Number of Operable OPRM Cells for Option Ill Stability at Brunswick 1 and 2," June 8, 2001.
| |
| Brunswick Unit 1 B 3.3.1.1-43 Brunwic Unt No. 71 I 1 3..1.-43Revision
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SIR 3.3.1.2.1 and SIR 3.3.1.2.3 (continued)
| |
| REQUIREMENTS Th-requen...... f ..........; 12 hourz for SR 3.3.1.2.1 .'c b.ased on SInsert 2 MODESr- 3 an 4, .... iit chan;.. ... es rc.. no ..... ed the.rcfore, the 12* hor... ,-.r...a..d ucc to 2.4÷ohour r
| |
| ... *SR 3.3.1.2.3.. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE.
| |
| ALTERATIONS are, in fact, OPERABLE. Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM.
| |
| [ Insert 1 Th...12..hour Frequncy is b....d up.on opcr....ng cxpcrncnc and steps tocsr ta h SRMs cqurdb h C r ntcpo q 1udfa r ,t * ~.*.÷-.. ,, ;*
| |
| * 1 'f** *+*
| |
| SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate with the detector inserted to the normal operating level, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.1.2-6 No. 31 I BrunwickUnit1 B.3.12-6Revision
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued)
| |
| REQUIREMENTS To accomplish this, the SR is modified by Note 1 that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. In addition, Note 2 states that this requirement does not have to be met during a core spiral offload. A core spiral offload encompasses offloading a cell on the edge of a continuous fueled region (the core cell can be offloaded in any sequence). If the core is being unloaded in this manner, the various core configurations encountered will not be critical.
| |
| DInsert Thc Fr'quon. is, bascd. up.
| |
| , hann rcdund.ncy and othcr inforatio
| |
| .. a..ab.. in thc.. c....... room., a*nd cn tha
| |
| ... .... rcq...............
| |
| ., .. .urc rc 12 hours to 24 hours.
| |
| S_.R 3.3.__1.2._.5 an._d S__RR3.3.1.2.6 f-[Insert 1 Performance of a CHANNEL FUNCTIONAL TEST demonstrate* the associated channel will function properly. SR 3.3.1.2.5 is requir d in MODE 5; and thc 7 day Frcqu..cncy ensures that the channels a e OPERABLE while core reactivity changes could be in progress.*-N functioning* b..,w,,n CH*LAMNl" NEL FUNCtTIONikAL TES*TS.
| |
| SR 3.3.1.2.6 is required to be met in MODE 2 with IRMs on Range 2 or below, and in"MODES 3 and 4. Sincc corc rcacti,,;t, chn,,,do not L (continued)
| |
| Insert 1I Brunswick Unit 1 ...-
| |
| B 3.3.1.2-7 eiinN.3 No. 31 I Revision
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued)
| |
| REQUIREMENTS The Note to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the .Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be performed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 3--a Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
| |
| SR 3.3.1.2.7 Insert I Performance of a CHANNEL CALIBRATION at a ,,,qunt of..24 verifies the performance of the SRM detectors and associated circuitry.
| |
| Thrmun' cnidr th ln oniin cuic opror h ct or comp............. The neutron detectors are excluded from the CHANNEL CALIBRATION (Note 1) because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified for a fixed useful life.
| |
| Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below. The allowance to enter the Applicability with the 24r1~hFrequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs (continued)
| |
| Brunswick Unit 1 B 3.3.1.2-8 BrunwickUnitI No. 31 I B.3.12-8Revision
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 (continued)
| |
| REQUIREMENTS consistent with the assumptions of the current plant specific setpoint methodology.
| |
| and ! 0). iucncyofl 18 day is blso on rcliai lltly anallyses (Rcfzl. 8, 9' SInsert I SR 3.3.:
| |
| 2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by selecting a control rod not in compliance with the prescribed sequence and verifying proper annunciation of the selection error, and by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour after any control rod is withdrawn in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until 1 hour after THERMAL POWER is
| |
| * 8.75% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is < 8.75% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the §2-day Frequency is not met per SR 3.0.2.
| |
| The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the I Insert 2 usu,,lly.pass .t-cS,,u,',llanc's w.,hen p'rforlnmod, ,at4thck 92 day"Frcqucncy SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in Table 3.3.2.1-1, one corresponding to each specific power range. The purpose of this SR is to assure that for each RBM power range, the RBM flux trip rod block outputs are enabled (not bypassed) and that the RBM flux trip setpoint being applied is equal to or more conservative than the specified (continued)
| |
| Brunswick Unit 1 ...- 1ReiinN.3 B 3.3.2.1-11 Revision No. 31 I
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.4 (continued)
| |
| REQUIREMENTS SR 3.3.2.1 .4.c is satisfied if, for an APRM Simulated Thermal Power level the high power level setpoint Allowable Value defined in the COLR, the RBM flux trip rod block outputs are not bypassed and the RBM flux trip setpoint being applied is less than or equal to the high trip setpoint Allowable Value defined in the COLR.
| |
| SR 3.3.2.1.5
| |
| * lInsert l1 The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be verified periodically to be
| |
| > 8.75% RTP. Ifthe RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not consieiredl inoperable Thc,- Frq ...... c baod on,-,t.h, trip, sctpoint A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch--Shutdown Position Function to ensure that the channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch--Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.
| |
| As noted in the SR, the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links.
| |
| This allows entry into MODES 3 and 4 if the 24 c-.tiR41 Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.2.1-13 Brunwic Unt No. 31 I I 3..2.-13Revision
| |
| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.2.1.6 (continued)
| |
| IInsert 1I J Survi,;llanc undcr ths conditionn that apply, du ring a plant"ouag ÷*a-lnd the pote-ntial for an u npln~neda tlra"nsie,,-nt i'f ther S r veilianca wero -pe4eR~e4 SR 3.3.2.1.7 A CHANNEL CALIBRATION is a complete check of the instrument loop.
| |
| and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The CHANNEL CALIBRATION may be performed electronically.
| |
| As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.8.
| |
| [ Inert The Fequenc.... is bae upon.the assumption of a 21 month calibrationn intra
| |
| ... i;n the,d4ot,-ermi*naton;* of the magnitude of equimen drift... i4;n the SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.
| |
| (continued)
| |
| Brunswick Unit I1 ...-
| |
| B 3.3.2.1-14 4ReiinN.3 Revision No. 31 I
| |
| | |
| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS B.1 (continued) 4 hour Completion Time provided in LCO 3.2.2 and LCO 3.2.3 for Required Action A.1, since this instrument's purpose is to preclude MCPR and LHGR violations.
| |
| 0.1 With the required channels not restored to OPERABLE status or placed in trip, THERMAL POWER must be reduced to < 23% RTP within 4 hours.
| |
| As discussed in the Applicability section of the Bases, operation below 23% RTP results in sufficient margin to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to protect fuel integrity during the feedwater controller failure, maximum demand event. The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to
| |
| < 23% RTP from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel REQUIREMENTS is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains feedwater and main turbine high water level trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption that 6 hours is the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the feedwater pump turbines and main turbine will trip when necessary.
| |
| SR 3.3.2.2.1 Performance of the CHANNEL CHECK oncc ovcry 2"1 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a (continued)
| |
| Brunswick Unit 1 B 3.3.2.2-5 No. 58 I BrunwickUnit1 B.3.22-5Revision
| |
| | |
| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE SR 3.3.2.2.1 (continued)
| |
| REQUIREMENTS similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or 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 plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limits.
| |
| Tho Frcgucncy ic baccd on oporating oxporionco that dcmonstratcc I Li i,...ii ii iL;i idii*iiL.; iL, i,..il;.;, . *l*,... *,,,.,.[3;-*,..F.. bLiiJi.;iL.;iiiL.;i ii.t, IL.;LL, i*iiii,..ii.
| |
| i ilL; ",.,..iq.;':* .......
| |
| SR 3.3.2.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| LInert SR 3.3.2.2.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the feedwater and main turbine valves is (continued'/
| |
| Brunswick Unit 1 B 3.3.2.2-6 BrunwickUnit1 No. 31 I B.3.22-6Revision
| |
| | |
| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE SR 3.3.2.2.3 (continued) Iinsert1i_
| |
| REQUIRMENTS included as part of this Surveillance and *verlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complet testing of the assumed safety function. Therefore, if a valve is incapab! of operating, the associated instrumentation would also be inoperable. Thc 24 month Frcgucncy ia b cDef4.%%
| |
| 3nI thia nIcc.d%
| |
| to p ,* Irfor tIII SIulr]1 llancc uI dc th11.aJ coldit.
| |
| J*IIiols thIat thc Sur.....nc wc......pcrformcd with thc..... c,. at .......
| |
| REFERENCES 1. UFSAR, Section 15.1.2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. GENE-770-06-1 -A, Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
| |
| Brunswick Unit 1 B 3.3.2.2-7 No. 31 I BrunwickUnitI B.3.22-7Revision
| |
| | |
| PAM Instrumentation B 3.3.3.1 BASES ACTIONS F._11 (continued)
| |
| Since alternate means of monitoring primary containment area radiation are available, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. These alternate means may be temporarily installed ifthe normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.
| |
| SURVEILLANCE As noted at the beginning of the SRs, the following SRs apply to each REQUIREMENTS PAM instrumentation Function in Table 3.3.3.1-1.
| |
| SR 3.3.3.1.1 Performance of the CHANNEL CHECK on..... 31.d.. -vcy ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or 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 plant instruments located throughout the plant.
| |
| Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its
| |
| [ Insrt I *,,* limit.
| |
| failr of..mor th. n onc ...... l Iof.. givcn Function in any 31 day intc....
| |
| ic4141*.
| |
| raIrc Th lJl.CHlA.NEL CHECKleuplcmctI cee1formlVI butJ ImIo.Irc
| |
| ......... ," .... r n nn i ouringn............na uce~3,.' OT ...
| |
| *-,,ilv il,.*,a,.* ,,,,.,,-,,,.4 .,u, LiiU iJiiJii;;;.;,* ,
| |
| (continued'*
| |
| Brunswick Unit 1 B 3.3.3.1-9 Brunwic Uni No. 40 I I 3.3A-9Revision
| |
| | |
| PAM Instrumentation B 3.3.3.1 BASES SURVEILLANCE SR 3.3.3.1.2 REQUIREMENTS (continued) Not Used.
| |
| SR 3.3.3.1.3 This SR requires a CHANNEL CALIBRATION to be performed.
| |
| CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies the channel responds to measured parameter with the necessary range and accuracy.
| |
| For Function 10, the CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.
| |
| LInert TI.,,* "IA *-I-, r"r*
| |
| i ii* *.l* IIIV, II,I, ilik;LiLiUiiL, V for CHANNELI ,C^ALIB^rlTrO of* PAM I t"u c tatl" of Table 3.3.3.1 1 is baccd on op ig cxpcricncc and uuuiG;i;;.;.;;~ W;i;r, tric Cr';;- ;rcTUc-,inc, c'yci- .
| |
| REFERENCES 1. Regulatory Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Revision 2, December 1980.
| |
| : 2. NRC Safety Evaluation Report, Conformance to Regulatory Guide 1.97, Rev. 2, Brunswick Steam Electric Plant, Units 1 and 2, May 14, 1985.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.3.3.1-10 Brunwic Unt No. 40 I I 3..3.-10Revision
| |
| | |
| Remote Shutdown Monitoring Instrumentation B 3.3.3.2 BASES (continued)
| |
| SURVEILLANCE SR 3.3.3.2.1 REQUIREMENTS Performance of the CHANNEL CHECK onqcc .... 31
| |
| .... enuesta 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 instrument channels could be an indication of excessive instrument drift in one of the channels or 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 plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized. For Function 2 of Table B 3.3.3.2-1, the CHANNEL CHECK requirement does not apply to the N017 instrument loop since this instrument loop has no displayed indication. The CHANNEL CHECK requirement does apply to the remaining instruments of Function 2.
| |
| I ne rt 1 ...... Frqu.... c .. b .... upon, plnt .... . in
| |
| ... CXpcric*ncc that SR 3.3.3.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy.
| |
| FfInsrt1-7.4.4. ,-JR4 UFSAR, wih Section h B.......
| |
| ... .... ... cyclc.
| |
| : 1. 50.36(c)(2)(ii).
| |
| REFERENCES CFR 10 REFERENCES 1. UFSAR, Section 7.4.4.
| |
| 2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 ...-
| |
| B3.3.3.2-4 eiinN.4 No. 41 Revision I
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.1 BASES ACTIONS D.1 and D.2 (continued) performs the intended function of the instrumentation (Required Action D.1). The allowed Completion Time of 6 hours is reasonable, based on operating experience, both to reach MODE 2 from full power conditions and to remove a recirculation pump from service in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel REQUIREMENTS is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.
| |
| SR 3.3.4.1.1 Performance of the CHANNEL CHECK oncc cvcr', 24-hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or 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 plant staff based on a combination of the channel instrument uncertainties, including indication (continued)
| |
| Brunswick Unit 1 B 3.3.4.1-7 No. 31 I BrunwickUnit1 B.3.41-7Revision
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.1 (continued)
| |
| REQUIREMENTS and readability. If a channei is outside the criteria, it may be an indication Inser 1
| |
| * that the instrument has drifted outside its limit.
| |
| ThcFrcucny i baod ,po.......n g expcrioncc that ..............
| |
| channPcl f*il'-rc is rars. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
| |
| SIR 3.3.4.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant Fi nert1-... specific setpoint methodology.
| |
| L ....... JTh"rk Feq o ,..,yof 92I da.y,, is ba -don thc rcl:abilitl-y Refife, li e-.**-*i]*
| |
| SR 3.3.4.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable ifthe trip setting is discovered to be less conservative than the Allowable Value specified in SIR 3.3.4.1.4. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant design analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than
| |
| [ Inset 1
| |
| * accounted for in the appropriate setpoint methodology.
| |
| Thc frcquncy of93asi.ac ntc ciblt nlsso SR 3.3.4.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL (continued)
| |
| Brunswick Unit 1 B 3.3.4.1-8
| |
| ...- eiinN.3 Revision No. 31 I
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.4 (continued)
| |
| REQUIREMENTS CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific fl Insr setpoint methodology.
| |
| Th.-, Frcucn, i* b.-, ,.- upon th..-
| |
| azooumpto ... f a,4mot.clbrto II*hf
| |
| ,,.,.., v,. , I h, *.,
| |
| ,,.., ,.. , i j*Li iii 4 ,i,, i II L~i ,,,. II i JSi *itii,.,,.,. tp ,...l,j ,ilJiii~ I i , , ,S ii, i i i i ,..,LS SR 3.3.4.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the design function. Therefore, if a breaker is incapable of operating, the associated instrument channel(s) would be inoperable.
| |
| [insert 1
| |
| * Th.........mo.nth Fr nc is b
| |
| .... v~uoin n.., i ocrom
| |
| -/o. i
| |
| *.A v,,..,ll v,.., *..llJ* l pot,1,;a1 for an unpann 4F l 'al',/ if.th
| |
| ... ............c wcrc pcrformcd wIIn mr r~a~mor am nnx~r Jnrrirtnn r~nrri"nr" nn~ ~v'mnn~rrnmcn mnn~r compn .ntw*ill ,,s,,II,. p -k .hc Survc'la' .... wh pr,,.,fcrm......
| |
| 24A month Frcquoncy.
| |
| REFERENCES 1. UFSAR Sections 5.4.1.2.4 and 7.6.1.3.1.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. GENE-770-06-1 -A, Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications, December 1992.
| |
| Brunswick Unit 1B3341-ReionN.1 B 3.3.4.1-9 Revision No. 31 I
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE and (b) for Functions other than 3.c provided the associated Function or REQUIREMENTS redundant Function maintains ECCS initiation capability. Upon (continued) completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 7) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.
| |
| * SR 3.3.5.1.1 Performance of the CHANNEL CHEC Koocc ..... ry24 hr ensrestha 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 instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited-ta 24-heurs; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| LInert chn,
| |
| ,, ,, a lur...
| |
| c ic,*,,
| |
| ra..
| |
| ,rc;,
| |
| ,''. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.5.1.2 and SR 3.3.5.1.6 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any (continued)
| |
| Brunswick Unit 1 B33512 B 3.3.5.1-29 eiinN.3 No. 31 I Revision
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.2 and SR 3.3.5.1.6 (continued)
| |
| REQUIREMENTS setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| SR 3.3.5.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analyses. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology.
| |
| ( Insert 1 Th... Fcu nc of.. .. 92.
| |
| .. a .. ba...... on c rclab
| |
| ....... an,ity iz of Re........ e7.
| |
| SR 3.3.5.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific fiIset 1 I* setpoint methodology.
| |
| Th rqcnyi acduo h
| |
| *;+,mtio o:*a.24*-*~f month cal+ibrto lnTiwrvui in Ti* ucTrrnInTion or tnc
| |
| * p-e;n+ 1se; (continued)
| |
| Brunswick Unit 1 B 3.3.5.1-30 Brunwic Unt No. 31 I 1 3..5.-30Revision
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic and simulated automatic operation for a specific channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this i Inerll Surveillance to complete testing of the assumed safety function.
| |
| REFERENCES 1. UFSAR, Section 5.2.
| |
| : 2. UFSAR, Section 6.3.
| |
| : 3. UFSAR, Chapter 15.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. (Deleted.)
| |
| : 6. UFSAR, Section 9.2.6.2.
| |
| : 7. NEDC-30936-P-A, BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation), Parts 1 and 2, December 1988.
| |
| Brunswick Unit 1 B 3.3.5.1-31 Brunwic Unt No. 58 I I 3..5.-31Revision
| |
| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel REQUIREMENTS is placed in an inoperable status solely for performance of required (continued) Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3, provided the associated Function maintains RCIC initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance.
| |
| That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.
| |
| SR 3.3.5.2.1 Performance of the CHANNEL CHECK once cvcry 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or 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 plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| LInert Th-rcqucn....... ic based upon .p..t.n cxp....... th..t dc..
| |
| , m....n tc*,
| |
| ch..nn.......u.. i,,z,.
| |
| r.r.. The CHANNEL CHECK suppements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.5.2-9 BrunwickUnitINo. 31 I B.3.52-9Revision
| |
| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant
| |
| [LIset 1 -,*, specific setpoint methodology.
| |
| -rh EF roQucncy ot U2 days ir hba^cd on tIhc rcl;abli-tly* anIl,*;z of SR 3.3.5.2.3 The calibration of trip units provides a check of the actual trip setpoints.
| |
| The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.2-1. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| I Insert 1 "The ml lrcoucnl *n OTV navoy 1D u I un,i LcIU IlnUlIItV Ref,efenCee-SR 3.3.5.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| I Insert 1 (continued'l Brunswick Unit 1 ...-
| |
| B 3.3.5.2-10 0ReiinN.3 Revision No. 31 I
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| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel and includes simulated automatic actuation of the channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to EI Iset1 *,, provide complete testing of the safety function.
| |
| r..fueling oua.
| |
| .. Opor...... cxpcricncc has dcmonstratod that thoco
| |
| ....... nc... wil '-'usualy pass teD
| |
| + ,hnpc.foermcd at thc 21t month,*
| |
| ,SR Froqucncy is conclu-ded to bo acceptablc from a reliabiliF,' standpoin~t.
| |
| REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| : 2. GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
| |
| Brunswick Unit 1 ...- 1Rvso B 3.3.5.2-11 o4 Revision No. 44 I
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| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES (continued)
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Primary REQUIREMENTS Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1.
| |
| The Surveillances are modified bY a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions with a design that provides only one channel per trip system and (b) for up to 6 hours for all other Functions provided the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 2 hour allowance for Functions with a design that provides only one channel per trip system or the 6 hour allowance for all other Functions, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 7 and 8) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 2 and 6 hour testing allowances do not significantly reduce the probability that the PCIVs will isolate the penetration flow path(s) when necessary.
| |
| SR 3.3.6.1.1 Performance of the CHANNEL CHECK oncc cvc,-; 21 ho'-rs ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the 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 plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.6.1-28 Brun wic Unt No. 73 I 1 3..6.-28Revision
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.1 (continued)
| |
| REQUIREMENTS
| |
| .. Thc Frcgucncy is bascd on opert.n exp.ri.nc. .. t h.t dom....tratos,
| |
| *h.. failurcis
| |
| ... rc... The- CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3~A.612 SR_3.3.6.1.5 and SR 3.3.6.1.9 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| I Insert 2 Thc. 92 day Froqucency of SR 3.3.6.1 .2 is basod on"th.c rcabiity anal.y,,si, doscribod in Roforcnccs 7 and 8. Tho 18"! day Frcqucncy of SR 3.3.6.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than that accounted for in the appropriate setpoint methodology.
| |
| I Insrt Thc Frsqucnc'; o.f 192 day*' is- based*,^ on thc Rcfcr....c. 71 an;d 8.
| |
| rciIaDIiity analysis or SR 3.3.6.1.4 and SR 3.3.6.1.6 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL (continued'*
| |
| Brunswick Unit 1 B 3.3.6.1-29 Brunwic Unt No. 73 I 1 3..6.-29Revision
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.4 and SR 3.3.6.1.6 (continued)
| |
| REQUIREMENTS CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| SR 3.3.6.1.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on PCI Vs in LCO 3.6.1.3 overlaps this Inser 1 *Surveillance to provideecompletetestingof, the..assumed safetyfunction.
| |
| this"Sur....a...c be perfo*, d,*'uring a unit outag".
| |
| ,'rmed* Op..rating' SR 3.3.6.1.8 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis.
| |
| Testing is performed only on channels where the assumed response time does not correspond to the diesel generator (DG) start time. For channels assumed to respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time (milliseconds) so as to assure adequate response without a specific measurement test (Ref. 9).
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.6.1-30 Brunwic Unt No. 73 I I 3..6.-30Revision
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.8 (continued)
| |
| REQUIREMENTS Note 1 to the Surveillance states that the radiation detectors are excluded from ISOLATION INSTRUMENTATION RESPONSE TIME testing. This Note is necessary because of the difficulty of generating an appropriate detector input signal and because the principles of detector operation virtually ensure an instantaneous response time. Response times for radiation detector channels shall be measured from detector output or the input of the first electronic component in the channel. In addition, Note 2 to the Surveillance states that the response time of the sensors for Functions l .a and 1 .c may be assumed to be the design sensor response time and therefore, are excluded from the ISOLATION INSTRUMENTATION RESPONSE TIME testing. This is allowed since the sensor response time is a small part of the overall ISOLATION INSTRUMENTATION RESPONSE TIME (Ref. 9).
| |
| !SOLATION INSTRUMENTATION RESPONSE TIME tcsts arc conductcd ona. 21 month, STAGGERE-Dn TrEST BASIS. Thc. 21A mont..h Frcqucncy i"
| |
| .... istcnt ,with÷ the. BNP .refueling c..c.o and is basod.. upon plant operatinqg experiec that show that random fiuoso nsrmntto REFERENCES 1. UFSAR, Section 6.3.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. UFSAR, Section 6.2.4.3.
| |
| : 6. UFSAR, Section 7.3.1.1.6.18.
| |
| : 7. NEDC-31 677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
| |
| : 8. NEDC-30851 P-A Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
| |
| : 9. NEDO-32291-A, System Analyses for Elimination of Selected Response Time Requirements, October 1995.
| |
| Brunswick Unit 1 B 3.3.6.1-31 Brun wic Unt No. 73 I I 3..6.-31Revision
| |
| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.1 REQUIREMENTS (continued) Performance of the CHANNEL CHECK oncc ewry 21 hou-rs ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive, instrument drift in one of the channels or 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 plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| channcl failuro i" ratc. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO.
| |
| SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| I Insert 1 I--*
| |
| Th FrqIc f9 a s is bascd on tho roliabilit,' analysis of SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| [ Inert
| |
| *asoa on tnoI Ir I IaHIIItV anivi H H Ol Rofocrncos 5 and 6.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.6.2-9 Brun wick UnitI No. 80 I B.3.62-9Revision
| |
| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.4 REQUIREMENTS (continued) A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| l Insert 1 *Th e-F*,
| |
| icncy is bascd on thc assumption of a 21 month calibration interm,.al in thc"dtr,',mirnatin,'f*, o.,f th," magnitude-
| |
| ' rarf o- equ,,ipontr -in ift v t;,FI hc SR 3.3.6.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on SCIDs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| fl Inr1 I I IJ l~l "'*Th,.e 21 mo.nth Fr-e,6vi 6 3
| |
| V isbasod on the need to pci 4G~Ri-fhis Sur.-e-illanco u-nder the conditions that apply during a plant outage and the
| |
| ,with tha reacto at.powe....r*.. Oprain e."periesce has demonstrated that tho-o compo".*n,',ent. 1ll.;I usua,,'ll,.. pass''* the Sr,,.-,ll,-n,-, w^hen perfo,'rmedatl-the* 21 month* Frequency.
| |
| REFERENCES 1. UFSAR, Section 15.6.4.
| |
| : 2. Not used.
| |
| : 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. NEDC-3 1677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
| |
| : 6. NEDC-30851P-A Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
| |
| Brunswick Unit 1 B 3.3.6.2-10 Brunwic Unt No. 80 I I 3..6.-10Revision
| |
| | |
| CREV System Instrumentation B 3.3.7.1 BASES ACTIONS B.._1 (continued)
| |
| The 1 hour Completion Time is intended to allow the operator time to place the CREV subsystem in operation. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, or for placing one CREV subsystem in operation.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel REQUIREMENTS is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains CREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel surveillance.
| |
| That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREV System will initiate when necessary.
| |
| SR 3.3.7.1.1 Performance of the CHANNEL CHECK onqcc cvcr; 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or 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 plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| (continued*
| |
| Brunswick Unit 1 B 3.3.7.1-5 No. 31 I BrunwickUnitI B.3.71-5Revision
| |
| | |
| CREV System Instrumentation B 3.3.7.1 BASES SURVEILLANCE SR 3.3.7.1.1 (continued)
| |
| REQUIREMENTS LInsert I channcl failu-ro is rarc. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with channels required by the LCO.
| |
| SR 3.3.7.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| Tlhe Froguoncy of 92 days is basod an thc roliability analyscs af SR 3.3.7.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| [Insert 1 Thc F"rcgnc is,bascd,.* upon* tho,, a....ptin~l of a 24 mat calib*~l-ration,,
| |
| Intra in* thI dIctc11minatIonII ofl~th maon1*II1u11dc of c,.ui,,1UlOmnt.dIft In thI SR 3.3.7.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.3, "Control Room Emergency Ventilation (CREV) System," overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.3.7.1-6 BrunwickUnitI No. 31 I B.3.71-6Revision
| |
| | |
| CREV System Instrumentation B 3.3.7.1 BASES SURVIELLANCE SR 3.3.7.1.4 (continued)
| |
| REQUIREMENTS
| |
| .... thc* Sur.o..a.. whon...
| |
| " porformed at tho 21 mon"th Frcgu--cncy.
| |
| REFERENCES 1. UFSAR, Section 6.4.4.1.
| |
| : 2. UFSAR, Section 15.7.1
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
| |
| Brunswick Unit 1 B 3.3.7.1-7 No. 36 I BrunwickUnitI B.3.71-7Revision
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES ACTIONS B.1. B.2, and B.3 (continued)
| |
| Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the Function not maintaining condenser vacuum pump isolation capability. The Function is considered to be maintaining condenser vacuum pump isolation capability when sufficient channels are OPERABLE or in trip such that the condenser vacuum pump isolation instruments will generate a trip signal from a valid Main Steam Line Radiation--High signal, and the condenser vacuum pumps will trip. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the condenser vacuum pump trip breakers to be OPERABLE.
| |
| SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel REQUIREMENTS is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the condenser vacuum pumps will isolate when necessary.
| |
| SR 3.3.7.2.1 Performance of the CHANNEL CHECK on.. c:'cry 24A or ensrestha 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 instrument channels could be an indication of excessive instrument drift in (continued)
| |
| Brunswick Unit 1 B 3.3.7.2-5 No. 31 I BrunwickUnitI B.3.72-5Revision
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES SR 3.3.7.2.1 (continued)
| |
| REQUIREMENTS one of the channels or 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 plant staff based on combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may b5e an indication that LiKIsr 1]-,,,, the instrument has drifted outside its limit.
| |
| Trhe Fr...u..nc. i's basd on thc CHuANNEL- CHECK Frcqucncy
| |
| ,rcqui........of..........n..........t.... The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
| |
| SR 3.3.7.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| FI Insr Jrqec Th f9 asi SR 3.3.7.2.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| I ~In.r1 Thc..Fcqu.n.. is ba..e. upon h asm tino an......monthe,,ir.....
| |
| intrva i th deerinaionofthcmagitdc f euimcn diftll* in! thcI wv*l
| |
| ~i 1
| |
| Brunswick Unit 1 B 3.3.7.2-6 BrunwickUnitI No. 31 I B.3.72-6Revision
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES SURVEILLANCE SR 3.3.7.2.3 (continued)
| |
| REQUIREMENTS For the purposes of this SR, background is the dose level experienced at 100% RATED THERMAL POWER with hydrogen water chemistry at the maximum injection rate. Under these conditions, an Allowable Value of
| |
| < 6 x background will ensure that General Design Criterion 19 limits of 10 CFR 50, Appendix A, will not be exceeded in the control room in the event of a CRDA.
| |
| SR 3.3.7.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers and actuation of the associated isolation valve are included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker is incapable of operating or the isolation valve is incapable of actuating, the instrument channel would be inoperable.
| |
| [ nert 1 ISu1vci llJlllanc uIc 1th-calj cnItin that*.J appl dur%*ingapln otg.adh potontiallql%,
| |
| for un,B lann d t=ra,,
| |
| i Ian icnU, tJLi ,.4 the
| |
| .,.#,lf *l lu~
| |
| il*nc woroi*,4 i pcrformcdil*Iil with the roactor at powor.
| |
| REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| : 2. NEDC-30851P-A, Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
| |
| Brunswick Unit 1 B 3.3.7.2-7 BrunwickUnitINo. 31 I B.3.72-7Revision
| |
| | |
| LOP Instrumentation B 3.3.8.1 BASES ACTIONS B..1.
| |
| (continued)
| |
| If any Required Action and associated Completion Time are not met, the associated Function is not capable of performing the intended function.
| |
| Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions for the inoperable DG(s).
| |
| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.8.1-1.
| |
| The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Function maintains initiation capability for three DGs; and (b) for Function 2, the associate Function maintains DG initiation capability. For Function 1, the loss of function for one DG for this short period is appropriate since only three of four DGs are required to start within the required times and because there is no appreciable impact on risk. Also, upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
| |
| SR 3.3.8.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the I Inset 1
| |
| * current plant specific setpoint methodology.
| |
| to c-hane
| |
| ... ER
| |
| '*DIL'A*IITVand drif-t, w^hich domonstrates th-at failur.. of (continued)
| |
| Brunswick Unit 1 B 3.3.8.1-5 Brun wick UnitI No. 78 I B.3.81-5Revision
| |
| | |
| LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE SR 3.3.8.1.2 and SR 3.3.8.1.3 REQUIREMENTS (continued)
| |
| A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology..
| |
| Tho Froqucncics of SR 3.3.8.1.2 and SR 3.3.8.1.3 arc based u-pon the SR 3.3.8.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed in LCO 3.8.1 and LCO 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions.
| |
| no;÷ 21 mont. I-re "ue.....---s or"r÷n"
| |
| -a.e tv-icnedrtonF ph-,or th...is-÷,, h*
| |
| Su-.**'*-,.,r-,,illance,*;;1 under*- the., codton htapuing,'.;l a-,*.*.-*
| |
| pln outage andth potetia fo nunlne........ transient,if...the..
| |
| Sr.,eillance* weep. fre h recoiih tuoe.Oertn xerec a demonstrted ihei REFERENCES 1. UFSAR, Section 6.3.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.3.8.1-6 Brun wick UnitI No. 78 I B.3.81-6Revision
| |
| | |
| RPS Electric Power Monitoring B 3.3.8.2 BASES (continued)
| |
| SURVEILLANCE SIR 3.3.8.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be performed while the plant is in a condition in which the loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. r-._*
| |
| The.1 day
| |
| . ,,.,nc... nd
| |
| .. thc Note in the Surveillance are based on guidance provided in Generic Letter 91-09 (Ref. 3). BN,.,.P*h,.*...,,,.*,.",,.
| |
| SR 3.3.8.2.2 and SR 3.3.8.2.3 Isr CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific FIiInetT1.-2.. setpoint methodology.
| |
| SR 3.3.8.2.4 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system will automatically trip open the associated power monitoring assembly. Only one signal per power monitoring assembly is required to (continued)
| |
| Brunswick Unit 1 B 3.3.8.2-6 BrunwickUnit1 No. 31 I B.3.82-6Revision
| |
| | |
| RPS Electric Power Monitoring B 3.3.8.2 BASES SURVEILLANCE SR 3.3.8.2.4 (continued)
| |
| REQUIREMENTS be tested. This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function. The system functional test of the Class 1E circuit breakers is included as part of this test to provide complete testing of the safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be inoperable.
| |
| [ Inert
| |
| * I*S.1 Il1.JIlLI**I
| |
| *l I*nt 1.191.41.411cy is bascd on tho nood to pc~orm this unlanned tr.... ....... t if.thc ............. werc per...rm.d....th.th. rcactor at whel*n
| |
| * at th.. 24 l÷t....perorm.... monrth
| |
| *'cqueney.
| |
| REFERENCES 1. UFSAR, Section 7.2.1.1.1.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. NRC Generic Letter 91-09, Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System.
| |
| Brunswick Unit 1 B 3.3.8.2-7 No. 31 I BrunwickUnitI B.3.82-7Revision
| |
| | |
| Recirculation Loops Operating B 3.4.1 BASES (continued)
| |
| SURVEILLANCE SR 3.4.1.1 REQU IREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 75% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can, therefore, be allowed when core flow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
| |
| The mismatch is measured in terms of the percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation.. ,-he REFERENCES 1. UFSAR, Section 5.4.1.3.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. 10 CFR 50.36(c)(2)(ii). I Brunswick Unit 1 ..-
| |
| B 3.4.1-5 eiinN.7 No. 77 Revision I
| |
| | |
| Jet Pumps B 3.4.2 BASES (continued)
| |
| SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This SR is designed to detect significant degradation in jet pump performance that precedes jet pump failure (Ref. 3). This SR is only required to be performed when the ioop has forced recirculation flow since surveillance checks and measurements can only be performed during jet pump operation. The jet pump failure of concern is a complete mixer displacement due to jet pump beam failure. Jet pump plugging is also of concern since it adds flow resistance to the recirculation loop.
| |
| Significant degradation is indicated if the specified criteria confirm unacceptable deviations from established patterns or relationships. The allowable deviations from the established patterns have been developed based on the variations experienced at plants during normal operation and with jet pump assembly failures (Ref. 3). Each recirculation loop must satisfy one of the performance criteria provided. Since refueling activities (fuel assembly replacement or shuffle, as well as any modifications to fuel support orifice size or core plate bypass flow) can affect the relationship between core flow, jet pump flow, and recirculation loop flow, these relationships may need to be re-established each cycle.
| |
| Similarly, initial entry into extended single loop operation may also require establishment of these relationships. During the initial weeks of operation under such conditions, while base-lining new "established patterns",
| |
| engineering judgement of the dai4y Surveillance results is used to detect significant abnormalities which could indicate a jet pump failure.
| |
| The recirculation pump speed operating characteristics (pump flow and loop flow versus pump speed) are determined by the flow resistance from the loop suction through the jet pump nozzles. A change in the relationship may indicate a plug, flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle. For this criterion, the pump flow and loop flow versus pump speed relationship must be verified. The generator speed associated with the recirculation pump motor-generator set may be used to measure recirculation pump speed.
| |
| (continued)
| |
| Brunswick Unit 1B342-ReionN.1 B 3.4.2-3 Revision No. 31
| |
| | |
| Jet Pumps B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 (continued)
| |
| REQUIREMENTS Individual let pumps in a recirculation loop normally do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers. The jet pump diffuser to lower plenum differential pressure pattern or relationship of one jet pump to the loop differential pressure ratio is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps.
| |
| The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 3).
| |
| Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in Reference 3.
| |
| T...21 hour Frcqucncy ha...bcc. h... b"""oper..ting to ...
| |
| This SR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation.
| |
| Note 2 allows this SR not to be performed when THERMAL POWER is
| |
| < 25% of RTP. During low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data.
| |
| REFERENCES 1. UFSAR, Section 6.3.3.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| : 3. GE Service Information Letter No. 330, June 9, 1980.
| |
| Brunswick Unit 1 B 3.4.2-4 Brunwic Uni No. 31 I I 3.42-4Revision
| |
| | |
| S RVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)
| |
| REQUIREMENTS the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow.
| |
| Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the required pressure is achieved to perform this test. Adequate pressure at which this test is to be performed, to avoid damaging the valve, is 945 psig. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.
| |
| Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure is adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is
| |
| [ Iser -1 considered OPERABLE.
| |
| Thc 21 month Froquency was doveloped bascd an the SRV tcsts dcmo.t.... t..,,
| |
| hat thcs. compnconts will u.u.ll. pa" thc, e ..... l ...
| |
| .hen.perfor..d..t.te.24.month c...n..
| |
| ... Thecfrc th...Frequ..ncy wa*S conc-lud-, to,,,bc- a,-cc,,,,+ta,,bl,,o.fromn- ,.ar,-,Iabill;t, st,,-nd-point.
| |
| REFERENCES 1. UFSAR, Section 5.2.2.2.
| |
| : 2. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Supplement 1, March 1996.
| |
| : 3. UFSAR, Chapter 15.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| *. AeSMEi-"dc-,fo.,,,r Opcratin and aInc,-nan-.,-co,,,-,.. o,,f Nuclea*lr Powe*r Brunswick Unit 1 B 3.4.3-4 Brunwic Uni No. 59 I I 3.43-4Revision
| |
| | |
| RCS Operational LEAKAGE B 3.4.4 BASES (continued)
| |
| SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate of the drywell and equipment drain sumps are monitored to determine actual LEAKAGE rates; however, any method may be used to quantify LEAKAGE within the guidelines of Reference 7. In conjunction with lrsan tc
| |
| "-4Insert 1I REFERENCES 1. 10 CFR 50.2
| |
| : 2. 10 CFR 50.55a(c).
| |
| : 3. UFSAR, Chapter 5.0.
| |
| : 4. GEAP-5620, Failure Behavior in ASTM A106B Pipes Containing Axial Through--Wall Flaws, April 1968.
| |
| : 5. NUREG-75/067, Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactors, October 1975.
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| : 7. Regulatory Guide 1.45, May 1973.
| |
| Brunswick Unit 1 B 3.4.4-5 Brunwic Uni No. 31 I 1 3.44-5Revision
| |
| | |
| RCS Leakage Detection Instrumentation B 3.4.5 BASES (continued)
| |
| SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the required primary containment atmosphere radioactivity monitoring system. The 1 nsr1l--* check gives reasonable confidence that the channel is operating properly.
| |
| SR 3.4.5.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in the desired manner. The test also verifies, for the radioactivity monitoring channels only, the required alarm function of the instrument string. A source check along with a channel check will be used to determine the relative accuracy of the instrument. Failure of the source check not attributed to an instrument indication problem (e.g., problem with source check mechanism and not the detector), would not immediately result in instrument inoperability. Tc,Fre..u..nc. of 31 day... con sid rs instrum cnt reliability, and, op....n
| |
| , xp...c... ha...chown
| |
| .... pro..... for dotocting SR 3.4.5.3 " .nsr1]
| |
| This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. Th-rcqucnc...... of 21 month+s,, is consistent 'with* th, REFERENCES 1. UFSAR, Section 5.2.5.
| |
| : 2. Regulatory Guide 1.45, Revision 0, "Reactor Coolant Pressure Boundary Leakage Detections Systems," May 1973.
| |
| : 3. GEAP-5620, Failure Behavior in ASTM A106B Pipes Containing Axial Through--Wall Flaws, April 1968.
| |
| : 4. NUREG-75/067, Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping in Boiling Water Reactors, October 1975.
| |
| : 5. UFSAR, Section 5.2.5.2.2.
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.4.5-5 Bruswik Uit B.4.-5Revision No. 76 I
| |
| | |
| RCS Specific Activity B 3.4.6 BASES ACTIONS B.1. B.2.1. B.2.2.1, and B.2.2.2 (continued)
| |
| If the DOSE EQUIVALENT 1-131 cannot be restored to
| |
| * 0.2 pCi/gm within 48 hours, or if at any time it is > 4.0 IpCi/gm, it must be determined at least once every 4 hours and all the main steam lines must be isolated within 12 hours. Isolating the main steam lines precludes the possibility of releasing radioactive material to the environment in an amount that is more than a small fraction of the requirements of 10 CFR 50.67 during a postulated MSLB accident.
| |
| Alternatively, the plant can be placed in MODE 3 within 12 hours and in MODE 4 within 36 hours. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads). In MODE 4, the requirements of the LCO are no longer applicable.
| |
| The Completion Time of once every 4 hours is the time needed to take and analyze a sample. The 12 hour Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing the unit in MODES 3 and 4 are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation* . . . ,,, ... . . .,... .. . ...... n This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.
| |
| REFERENCES 1. 10 CFR 50.67.
| |
| : 2. UFSAR, Section 15.6.3.
| |
| : 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, dated September 1995.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. 10 CFR 50, Appendix A, GDC 19.
| |
| Brunswick Unit 1 B 3*4.6-3 Brunwic Uni No. 41 I I 3.46-3Revision
| |
| | |
| RHR Shutdown Cooling System--Hot Shutdown B 3.4.7 BASES (continued)
| |
| SURVEILLANCE SIR 3.4.7.1 [Insert1I REQUIREMENTS This Surveillance verifies that one required IR, R shutdown cooling subsystem or recirculation pump is in operation *nd circulating reactor coolant. The required flow rate is deter~mined by. e flow rate necessary to provide sufficient decay heat removal capability. Thc Frcquconcy cf 12 hou.,m ic ; ufcc. nti.n;**4 vio of;*otho vi.
| |
| ,, ua
| |
| , nd aud,,i blo,-;.-I
| |
| ;,i;dic -,
| |
| .... !1_1_ 1_
| |
| * I LI .. ... *
| |
| .............. g
| |
| , I,, I, ( ......
| |
| nc ..... ato., ,or .. on. torln,. inc, ... . . .i c
| |
| inqC c...,,o, in....
| |
| This Surveillance is modified by a Note allowing sufficient time to align the IRHIR System for shutdown cooling operation after the pressure setpoint that isolates the shutdown cooling mode of the IRHIR System is reset, or for placing a recirculation pump in operation. The Note takes exception to the requirements of the Surveillance being met (i.e., forced coolant circulation is not required for this initial 2 hour period), which also allows entry into the Applicability of this Specification in accordance with SIR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
| |
| REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B3.4.7-5 Brunwic Uni No. 41 I I 3.47-5Revision
| |
| | |
| RHR Shutdown Cooling System--Cold Shutdown B 3.4.8 BASES (continued)
| |
| SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance verifies that one required RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.jT he Frcquoncy of rrs su ffi.dent, In vArnrw of other v*sua nndj,~auntie indncsnon
| |
| !2) hors, Insert 1I REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3*4.8-5 Brunwic Uni No. 31 I 1 348-5Revision
| |
| | |
| RCS P/T Limits B 3.4.9 BASES (continued)
| |
| SURVEILLANCE SR 3.4.9.1 and SR 3.4.9.2 REQUIREMENTS Verification that operation is within limits is required cvcry 30 minu'tcs when RCS pressure and temperature conditions are undergoing planned I Insert 2 *changes. This Frc........ is c..... , rc rcao...... in^
| |
| ;v.., .. f. t. c..ntrol Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
| |
| SR 3.4.9.1 is modified by a Note that requires the Surveillance to be performed only during system heatup and cooldown operations.
| |
| SR 3.4.9.2 is modified by a Note that requires the Surveillance to be performed only during inservice leakage and hydrostatic testing.
| |
| SR 3.4.9.3 A separate limit is used when the reactor is approaching criticality.
| |
| Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.
| |
| Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
| |
| SR 3.4.9.4 and SR 3.4.9.5 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
| |
| (continued)
| |
| Brunswick Unit 1 B 3.4.9-7 Brunwic Uni No. 31 I I 3.49-7Revision
| |
| | |
| RCS P/T" Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.4 and SR 3.4.9.5 (continued)
| |
| REQUIREMENTS Performing the Surveillance within 30 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
| |
| An acceptable means of demonstrating compliance with the differential temperature requirement of SR 3.4.9.4 is to compare the temperature of the reactor coolant in the dome to the bottom head drain temperature.
| |
| As specified in procedures, an acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.5 is to compare the temperatures of the operating recirculation loop and the idle loop.
| |
| SR 3.4.9.4 and SR 3.4.9.5 are modified by a Note that requires the Surveillance to be met only in MODES 1, 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required. The Note also states the SR is only required to be met during recirculation pump startup, since this is when the stresses occur.
| |
| SR 3.4.9.6, SR 3.4.9.7, and SR 3.4.9.8 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
| |
| The flange temperatures must be verified to be above the limits
| |
| ...... nui.es before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature < 80°F, 30-wir,'tc checks of the flange temperatures are required because of the reduced margin to the limits.
| |
| When in MODE 4 with RCS temperature < 100°F, monitoring of the flange temperature is required cv'ery-A2-heurs to ensure the temperature is within the specified limits.
| |
| (continued')
| |
| Brunswick Unit 1 B3.4.9-8 Brun wic Uni No. 31 I I 3.49-8Revision
| |
| | |
| RCS P/T Limits B 3.4.9 BASES SR 3.4.9.6, SR 3.4.9.7. and SR 3.4.9.8 (continued)
| |
| REQUIREMENTS Thc 30mnut;,,o Fr..........rct tho ur....c. of maintaining tho Insert 2 limit .. ould" bc c...dc..., Tho 12 hour Frcguoncy i; r........ acd o,*,
| |
| n SR 3.4.9.6 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs.
| |
| SR 3.4.9.7 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature is < 80 0 F in MODE 4.
| |
| SR 3.4.9.8 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature is __1000 F in MODE 4. The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the specified limits.
| |
| REFERENCES 1. Calculation 0B21-1029, "Instrument Uncertainty for RCS Pressure/Temperature Limits Curve," Revision 0.
| |
| : 2. 10 CFR 50, Appendix G.
| |
| : 3. 1989 Edition of the ASME Code, Section Xl, Appendix G.
| |
| : 4. ASME Code Case N-640. "Alternate References Fracture Toughness for Development of P-T Limit Curves Section Xl.
| |
| Division 1 ."
| |
| : 5. EPRI Report TR-1003346, BWRVIP-86-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan, October 2002.
| |
| : 6. 10 CFR 50, Appendix H.
| |
| : 7. Regulatory Guide 1.99, Revision 2, May 1988.
| |
| : 8. Calculation 0B1 1-0005, "Development of RPV Pressure-Temperature Curves For BNP Units 1 and 2 For Up To I 32 EFPY of Plant Operation," Revision 1.
| |
| : 9. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
| |
| : 10. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1B349-ReionN.8 B 3.4.9-9 Revision No. 38 I
| |
| | |
| Reactor Steam Dome Pressure B 3.4.10 BASES APPLICABILITY In MODES (continued) shut down. 3,In4,these and 5, the limitthe MODES, is not applicable reactor because pressure is well the reactor below the is required limit, and no anticipated events will challenge the overpressure limits.
| |
| ACTIONS A.1*
| |
| With the reactor steam dome pressure greater than the limit, prompt action should be ta'ken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is reasonable considering the importance of maintaining the pressure within limits. This Completion Time also ensures that the probability of an accident occurring while pressure is greater than the limit is minimized.
| |
| B.1 If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours 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 3.4.10.1 REQUIREMENTS Verification that reactor steam dome pressure is _ 1045 psig ensures that the initial conditions of the vessel overpressure protection analysis are REFERENCES 1. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, dated September 1995.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B341-B 3.4.10-2 Revision eiinN.3 No. 31
| |
| | |
| ECCS--Operating B 3.5.1 BASES ACTIONS J.__
| |
| (continued)
| |
| When multiple ECCS subsystems are inoperable, as stated in Condition J, the plant is in a condition outside of the accident analyses.
| |
| Therefore, LCO 3.0.3 must be entered immediately.
| |
| SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping of each ECCS has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS subsystems, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This SR also prevents water hammer in the piping following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. The-g3!-dey SR 3.5.1.2 Isr Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these are verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition to the accident position in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
| |
| (continued'*
| |
| Brunswick Unit 1 B 3.5.1-10 Brun wick Uni No. 31 I I B3.5.-10Revision
| |
| | |
| ECCS--Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 (continued)
| |
| REQUIREMENTS Program requi-'rements fo.... rmrr;g valve tcsting at Icast"once......
| |
| IInsert 1I th,.-,
| |
| 92o days+O.*.,-,The Frqec ... of.31 daytsO fute -.-
| |
| is .u.t.... because','.
| |
| ar-e.;*.-
| |
| operated.,-lh under~hproedr al-,*- control-,I-. and, becau-so:,.,r.
| |
| Timp;-ropr,,-, ,,.al,, e -,-
| |
| poition t....ypically onl..'
| |
| .. y affect....
| |
| a n ..o .. ,usstm Thi....... Freqenc ha-s..
| |
| In MODE 3 with reactor steam dome pressure less than the RHR shutdown cooling isolation pressure, the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, this SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes the period when the required RHR pump is not operating and the period when the system is being realigned to or from the RHR shutdown cooling mode. At low reactor pressure and with a low decay heat load associated with operation in MODE 3 with reactor steam dome pressure less than the RHR shutdown cooling isolation pressure, a reduced complement of low pressure ECCS subsystems should provide the required core cooling in the unlikely event of a LOCA, thereby, allowing operation of the shutdown cooling mode of the RHR System, when necessary.
| |
| SR 3.5.1.3 Verification every 4! d~ys that ADS pneumatic supply header pressure is
| |
| _ 95 psig ensures adequate pneumatic pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least three valve actuations can occur with the drywell at 70% of design pressure. The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of > 95 psig is provided by the non-interruptible Reactor (continued)
| |
| Brunswick Unit 1 ..- 1ReiinN.3 B 3.5.1-11 Revision No. 31 I
| |
| | |
| ECCS--Operating B 3.5.1 BASES SURVEILLANCE SIR 3.5.'1.3 (continued) - Insert1I REQUIREMENTS Instrument Air System, the Pneumatic Nitrog n System, or the Nitrogen Backup System. This SR may be satisfied *,verifying the absence of all associated pneumatic low pressure alarms.V'- ho 31 day Frcgucncy takos irno cnnilnnr'illnn qnmlniriifnziiv Cnfrnl#i oveor nnnrefinn 0T Tnc pncu~malucr'
| |
| ..... ...m+ d'larms for
| |
| * low ..ir and ...t..gc, pressurc.
| |
| SR 3.5.1.4 Verification every 3*4 .... y that the RHR System cross tie valve is locked closed ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem. If the RHR System cross tie valve is open, both LPCI subsystems must be considered inoperable. Thc 31 day, Frc.......,ncy ,
| |
| Cycling the recirculation pump discharge and bypass valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will close when required. Upon initiation of an automatic LPCl subsystem injection signal, these valves are required to be closed to ensure full LPCl subsystem flow injection in the RPV.
| |
| De-energizing the valves in the closed position will also ensure the proper flow path for the LPCI subsystem. Acceptable methods of de-energizing a valve include de-energizing breaker control power, racking out the breaker or removing the breaker.
| |
| The specified Frequency is once each reactor startup before THERMAL POWER is > 25% RTP. However, this SR is modified by a Note that states the Surveillance is only required to be performed ifthe last performance was more than 31 days ago. Verification prior to or during each reactor startup prior to reaching > 25% RTP is an exception to the normal Inservice Testing Program generic valve cycling Frequency of (continued)
| |
| Brunswick Unit 1 ..- 2ReiinN.3 B 3.5.1-12 Revision No. 31 I
| |
| | |
| ECGS--Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 (continued)
| |
| REQUIREMENTS Therefore, SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes that state the Surveillances are not required to be performed until 48 hours after the reactor steam pressure is adequate to perform the test.
| |
| SR 3.5.1.89sbsdo h edt efomteSrelac ne h conirdeignfntions.ta pl proSurdrvingca Thust veiisthatup from aplatqutaged Opertemingtiexperieince hactudemonsrsiuated)th, thes uomptciiitonet illgi uoull Pass theSR nLC when paueromda the 2ystmsont Frequsency whicherit base desithed rencluing cactue.o oftherefore, therequencytitscosiergednto baoperabing fromnce aureliabltyc tnpoint.ru n cuainol the EPCS Subystems arwrqirdtoatut automatically o nRP o totrpaerforme thirnadeseignd funcios.qenthi SelanceP veiies thatr welthi anrequirhed syseuiiition ssatoaignall (atualsforesiomuae) the auTtomatic initiatsion logic of HaICS, andlevCl willnauslh orasupeio ytm subsysh wtems toeoerigate aEGS deIgnstuedincldngatation,"ooflp thes systemacth rougotitsecmpergenc operating sfteqasuene, sautomtic pumpctaruiadatutono.l automatic.. va.;*;nlve ,t requi..redfpoitins ...
| |
| their,, This Raloensuresthat theHPI ystmil auomtiall rstrt n n PVlowaelvl signl reeivd susequnt o anRPVhighwate leel tip ondthatnted Brunswick Unit 1B351-4RvsoN.4 B 3.5.1-14 Revision No. 44 I
| |
| | |
| ECOS--Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 (continued)
| |
| REQUIREMENTS This SR is modified by a Note that excludes vessel injection/spray during I the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
| |
| SR 3.5.1.10 The ADS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A sy~stem functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
| |
| solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SIR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete
| |
| [ Iset KJ ,, testing of the assumed safety function.
| |
| The. 21 month* Fr.......e. is based on tho need to+. pe."form thc potential.for.an unplanned tr-ansion.t if thc. SuR'eilla*nce were.. pc .... .d wi'th the rcactor at powcr. Opcrating cxporience has demonstrated that
| |
| ........ co pnot
| |
| .. ill...u.u.. ll.y pass th.. SR h .. pc.rormed at th 2mont Frequenc...., "whichis based"on +hcrefucling cycle. Therefore, the. Frequency.. is consdere to.... be. accep.
| |
| ... table"from a reliabilit' standpoint.
| |
| This SR is modified by a Note that excludes valve actuation since the valves are individually tested in accordance with SR 3.5.1.11. This also prevents an RPV pressure blowdown.
| |
| (continued)
| |
| Brunswick Unit 1 B B35113.5.1-15 eiinN.4 Revision No. 44 i
| |
| | |
| ECCS--Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 REQUIREMENTS (continued) A manual actuation of each required ADS valve is performed to verify that the valve and solenoid are functioning properly and that no blockage exists in the SRV discharge lines. This is demonstrated by the response of the turbine control or bypass valve; by a change in the measured flow; or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the required pressure is achieved to perform this SR. Adequate pressure at which this SR is to be performed, to avoid damaging the valve, is 945 psig. Reactor startup is allowed prior to performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure is adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance. SR 3.5.1.10 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
| |
| FI Insr L ...... *'1"The* Frequencyt of-*21 mont,.*hs i 3e on th
| |
| *noodl to pnnrforl the
| |
| *A
| |
| ~ur'.,L;;:uIIuu ur,~ur tric uur1uIuurI~
| |
| star'tup from a plant outage. prat..n. oxpn"ico has domonstrated that÷ those* components.*÷ will usua,-lly pass the..SR* .. d at t.he w.he porform SR 3.5.1.12 This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be the design instrumentation response time. Therefore, the instrumentation response time is excluded from the ECCS RESPONSE (continued)
| |
| Brunswick Unit 1 ..- 6ReiinN.3 B 3.5.1-16 Revision No. 31 I
| |
| | |
| ECCS--Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 (continued)
| |
| REQUIREMENTS TIME testing. This exception is allowed since the ECCS instrumentation response time is a small part of the ECCS RESPONSE TIME (e.g.,
| |
| sufficient margin exists in the emergency diesel generator start time when
| |
| [ Inset 1
| |
| * compared to the instrumentation response time) (Ref. 14).
| |
| REFERENCES 1. UFSAR, Section 6.3.2.2.3.
| |
| : 2. UFSAR, Section 6.3.2.2.4.
| |
| : 3. UFSAR, Section 6.3.2.2.1.
| |
| : 4. UFSAR, Section 6.3.2.2.2.
| |
| : 5. UFSAR, Section 15.2.
| |
| : 6. UFSAR, Section 15.6.
| |
| : 7. 10 CFR 50, Appendix K.
| |
| : 8. UFSAR, Section 6.3.3.
| |
| : 9. 10 CFR 50.46.
| |
| : 10. (Deleted.)
| |
| : 11. 10 CFR 50.36(c)(2)(ii).
| |
| : 12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
| |
| Recommended Interim Revisions to LCOs for ECCS Components, December 1, 1975.
| |
| : 13. UFSAR, Section 6.3.3.7.
| |
| : 14. NEDO-32291-A, System Analyses for the Elimination of Selected Response Time Testing Requirements, October 1995.
| |
| Brunswick Unit 1 B 3.5.1-17 BrunwickUniNo. 58 I I B3.5.-17Revision
| |
| | |
| ECCS--Shutdown B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 (continued)
| |
| REQUI REMENTS level is >_-31 inches or that the CS pump is aligned to take suction from the CST and the CST contains a total volume, which includes both usable and unusable volumes, of _>228,200 gallons of water, ensures that the CS System can supply at least 50,000 gallons of makeup water to the RPV. CS System air ingestion is expected to occur at the level which corresponds to a CST volume of 178,200 gallons. However, as noted, only one required CS subsystem may take credit for the CST option during OPDRVs. During OPDRVs, the volume in the CST may not provide adequate makeup if the RPV was completely drained. Therefore, only one CS subsystem is allowed to use the CST. This ensures the other required ECCS subsystem has adequate makeup volume.
| |
| The 12 hour Frccjucncy of thc~c ,., wac* do-v-e-,,,*l,-pod,- ,-,,-, de,;dr;i-opertinlo-cloxp..............
| |
| water... nc.. related F h to ........
| |
| .... r-a, pol....
| |
| .....12 hour* Frequenc ovlIc ndCSconid-ro
| |
| ,-# ...t.r to an.abnor.al suppression pool-o CST 'waterlc'-ol condition.
| |
| SR 3.5.2.3. SR 3.5.2.5. SR 3.5.2.6. and SR 3.5.2.7 The Bases provided for SR 3.5.1.1, SR 3.5.1.6, SR 3.5.1.9, and SR 3.5.1.12 are applicable to SR 3.5.2.3, SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7, respectively.
| |
| SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed (continued)
| |
| Brunswick Unit 1 B 3.5.2-5 Brunwic Uni I No. 31 I 3.52-5Revision
| |
| | |
| ECCS--Shutdown B 3.5.2 BASES S URVE ILLANCE SR 3.5.2.4 (continued)
| |
| REQU IREM ENTS to be in a nonaccident position provided the valve will automatically reposition to the accident position in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The-3!-day
| |
| .....u.... is.pp..p.... b-cauc.... thc.... o arc... oporatod under proccdural contro-+,-l and.. ,.c probability;+ of thcir boing mi-spo..itionod-' during this timc period is low.
| |
| In MODES 4 and 5, the RHR System may be required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, this SR is modified by a Note that allows one LPCI subsystem to be considered OPERABLE if it is capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. Alignment and operation for decay heat removal includes the period when the required RHR pump is not operating and the period when the system is being realigned to or from the RHR shutdown cooling mode. Because of the low pressure and low temperature conditions in MODES 4 and 5, sufficient time is available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery. This will ensure adequate core cooling if an inadvertent RPV draindown should occur.
| |
| REFERENCES 1. NEDO-20566A; General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K, Vols. 1, 2, and 3; September 1986.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit I B 3.5.2-6 Brunwic Uni I No. 31 I 3.52-6Revision
| |
| | |
| RCIC System B 3.5.3 BASES (continued)
| |
| SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge line of the RCIC System full of water ensures that the system will perform properly, injecting its full capacity into the reactor vessel upon demand. This SR will also prevent water hammer in the piping following an initiation signal. One acceptable method of ensuring the line is full is to vent at the high points. e-.-y System pipi.ng-, the procedural control-s go.erning sy...tem operation, and opertin c;" xperience. t* Isr tli SR 3.5.3.2 Verifying the correct alignment for manual, power operated, and automatic valves in the RCIC flow path provides assurance that the proper flow path exists for RCIC System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition to the accident position in the proper stroke time.
| |
| This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
| |
| This SR also includes the steam flow path for the turbine and the flow controller position.
| |
| J Insert 1 The 31..
| |
| " da- Frequ..ncy of his SR ... s derived... from, the. ..... i Testing
| |
| ,Program .....ie...t. for*"pe,"forming valve testing* at*least.once eve'P 92 days... The Frequency. of 31 days is fu'he ' justified, beca...s. the valves..
| |
| .r.oerte under.pocedura control....and. beas improper.....va...ve postin'pica.yr-*. affect only the* ROI System*,. Thi Frequency1..9 has (continued)
| |
| Brunswick Unit 1B3534ReionN.1 B3.5.3-4 Revision No. 31 J
| |
| | |
| RCIC System B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 (continued)
| |
| REQUI REMENTS inlet is OPERABLE, Note 1 to SR 3.5.3.3 requires the high pressure test to be performed with the turbine steam being supplied with reactor steam from the Main Steam System.
| |
| A 92 day, Frequenc... for SR 3 *.5 *;..3.icon.istc÷
| |
| . with thc. in.... *,.c Te-sting; Prog,,r.m requ,,,irements. The 2,4 mont-.h Frr,",'*,cqucn, for SR-.5.3.1 4 is- ba,,- on thc necd,* to perform tho Su~rv;llnc ,undcrcond;tion that e-xpcrienoc h.as demonstrated that÷ thcse. components ,wW,u.ually pa., thc-The RCIC System is required to actuate automatically in order to verify its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence; that is, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures the RCIC System will automatically restart on an RPV low water level signal received subsequent to an RPV high water level trip and that the suction is automatically transferred from the CST to the suppression pool on a CST low level signal. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.2, "Reactor Core Isolation Cooling (RCIC) System Instrumentation," overlaps this Surveillance to provide complete testing of the assumed design function.
| |
| [Insert While this Surveillance can be performed with the reactor at power, operating experience has demonstrated that the se components...will usuall',' pass the SR when performed at the 21 nnf *ftlrl *r*rll Enn* *nl*n 1*
| |
| based on tho refueling cycle. Therefore, the Frc be acceptable from a reliabilit',' standpoint.
| |
| (continued)
| |
| (continued)
| |
| Brunswick Unit 1 B 3.5.3-6 Brunwic Uni No. 31 I 1 3.53-6Revision
| |
| | |
| Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.1 (continued)
| |
| REQUIREMENTS (continued) leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of < 1.0 La. At < 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. The Frequency is required by the Primary Containment Leakage Rate Testing Program.
| |
| SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requi~rements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
| |
| SR 3.6.1.1.2 Maintaining the pressure suppression function of primary containment requires limiting the leakage from the drywell to the suppression chamber.
| |
| Thus, if an event were to occur that pressurized the drywell, the steam would be directed through the downcomers into the suppression pool.
| |
| This SR measures drywell to suppression chamber differential pressure during a 10 minute period to ensure that the leakage paths that would bypass the suppression pool (downcomers) are within allowable limits.
| |
| Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the differential pressure between the suppression chamber and the drywell does not decrease by more than 0.25 inch of water per minute over a 10 minute period. The,*-"ek"a'-'-e *'*
| |
| is poformcd,.. cvcry* 21 month.s..1. "Thc21 month.-* IFrcqucn.,y was. ,d1el1p,-
| |
| SIsethf1-REFERENCES 1. UFSAR, Section 6.2.
| |
| : 2. UFSAR, Section 15.6.
| |
| : 3. 10 CFR 50, Appendix J, Option B.
| |
| : 4. NEDC-33039P, Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Extended Power Uprate, August 2001.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.
| |
| : 7. Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.
| |
| (continued'*
| |
| Brunswick Unit 1 B 3.6.1.1-5 No. 57 I BrunwickUnitI B.6.11-5Revision
| |
| | |
| Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.2 REQUIREMENTS (continued) The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of the air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary 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 inner and outer door opening will not inadvertently occur.
| |
| inter~lockl mcchnismu islnot* i~nomlycaln onged whnnto rma tric ..... r.......to ....ng doo.. opcnng) this t ihetonl.......
| |
| .. ..to..
| |
| prformedlcycr 21.,; ,. monhF Thc 21*; motIFounc1sae o h wnco toll pcrlr this Sur*el la*nc unde hecnitonthtapl urn, REFERENCES 1. UFSAR, Section 3.8.2.4.3.2.
| |
| : 2. NEDC-33039P, Safety Analysis Report for Brunswick Units 1 and 2 Extended Power Uprate, August 2001.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. 10 CER 50, Appendix J, Option B.
| |
| : 5. NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.
| |
| : 6. Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.
| |
| : 7. NRC SER, Brunswick 1 &2 - Amendments No. 10 and 36 to Operating Licenses Revising Technical Specifications to Grant Exemptions from Specific Requirements of 10 CFR 50 Appendix J, dated November 8, 1977.
| |
| Brunswick Unit 1 B 3.6.1.2-8 Brunwic Unt 1B No. 46 I 36.12-8Revision
| |
| | |
| PCI Vs B 3.6.1.3 BASES ACTIONS F.1 and F.2 (continued) restore the valve(s) to OPERABLE status. This allows RHR shutdown cooling to remain in service while actions are being taken to restore the valve.
| |
| SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.
| |
| This SR does not require any testing or valve manipulation. Rather, it involves verification that those devices outside primary containment, and capable of being mispositioned, are in the correct position. S4iaee
| |
| %, r n,-,I,-÷;
| |
| ,-,1 ,n-* x I-I,n "nc'1 ,*L-*,, n m--,. r ,.,r v na.+p~,-,
| |
| ,, ,;,,-.ne m,,-,,-rI,,, n,--.n;,,- --'*,-r,I,-,, ;1 assurance that thc de-'ices arc in thc correct positions." Isr1 Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these devices, once they have been* verified to be in the proper position, is low.
| |
| A second Note has been included to clarify that PCI Vs that are open under administrative controls are not required to meet the SR during the time that the PCI Vs are open. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated. This SR does not apply to valves and blind flanges that are locked, sealed, or otherwise secured in the correct position, since these devices were verified to be in the correct position upon locking, sealing, or securing.
| |
| (continued)
| |
| Brunswick Unit 1 B3.6.1.3-9 BrunwickUnit1 No. 31 I B.6.13-9Revision
| |
| | |
| PCI Vs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.3 (continued)
| |
| REQUIREMENTS administrative controls, such as those that limit the shelf life of the explosiv charges~t, mutb olwd h 1da rgoc bcdo chrg oniuiy Insert 1/
| |
| SR 3.6.1.3.4 Verifying the isolation time of each power operated and each automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.5. The isolation time test ensures that each valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the requirements of the Inservice Testing Program.
| |
| SR 3.6.1.3.5 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA and transient analyses. This ensures that the calculated radiological consequences of these events remain within 10 CFR 50.67 limits. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.
| |
| SR 3.6.1.3.6 Automatic PCI Vs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. This SR includes verifying that each automatic PCIV in the Containment Atmosphere Dilution System flow path will actuate to its isolation position on the associated Group 2 and 6 primary containment isolation signals.
| |
| The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.1, "Primary Containment Isolation Instrumentation," overlaps this SR to provide (continued)
| |
| Brunswick Unit 1 B 3.6.1.3-11 Brunwic Unt No. 31 I I 3..1.-11Revision
| |
| | |
| PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.6 (continued) ,-[Isr1 REQUIREMENTS complete testing of the safety function. Thc 21 monqth Frogueney was coln ....... Ter fowandirupt the nrormal-'3 operti. of many.,critical. .,.
| |
| This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCVs) is OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break signal. This may be accomplished by cycling the EEC Vs through one complete cycle of full travel. The representative sample consists of an approximately equal number of EFCVs, such that each EFCV is tested at least once e.ery 10, years
| |
| ("'";"-i,,a,,, ,,. In addition, the EFCVs in the samples are representative of the various plant configurations, models, sizes, and operating environments.
| |
| This ensures that any potentially common problem with a specific type or application of EFCV is detected at the earliest possible time. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during a postulated instrument line break event. The 21t month Freqenc is..;
| |
| J Insert 1 the Srvoilanco ere prformed ,w,.ththe reacto atowr.Oprain Sur....illno w....he p..e,-Formed at-t.,-,h,. 2,1 mont,.,,,h Frequency.... Therefore, the-Frequen.y.was.conclude to*'be acceptable from^a..reliability *,,- standp*oin..-
| |
| has demonstrated that these
| |
| ................. riu rI~oII~
| |
| components are highly uperating miniicu reliable and experience that failures to isolate are very infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint.
| |
| (continued')
| |
| Brunswick Unit 1 B 3.6.1.3-12 Bruswik No. 46 I UitIB3.61.312Revision
| |
| | |
| Drywell Air Temperature B 3.6.1.4 BASES SURVEILLANCE SR 3.6.1.4.1 (continued)
| |
| REQUIREMENTS The following locations are monitored to obtain the drywell average temperature:
| |
| : a. Below 5 ft elevation;
| |
| : b. Between 10 fit and 23 ft elevation;
| |
| : c. Between 28 fit and 45 fit elevation;
| |
| : d. Between 70 ft and 80 fit elevation; and L Iset
| |
| * e. Above 90 fit elevation.
| |
| Th ll hour FlroqlIcncV of thlc SR is blsc onV opIratinglll oxperiene related, to, dr,... el avcrage air tcmperaturc variations and,, temperature theqcontrol room to aler-t the operator to an*'*abnormal ,*, dr'y-o-l airo; temperature.. condition.
| |
| REFERENCES 1. UFSAR, Section 6.2.
| |
| : 2. GE-NE-A22-00113-22-01, Brunswick Nuclear Plant Units 1 and 2, Extended Power Uprate - Task T0400 - Containment System Response, May 2001.
| |
| : 3. UFSAR, Section 6.2.1.1.1.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.6.1.4-3 BrunwickUnit1 B.6.14-3Revision No. 31
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.1 (continued)
| |
| REQUIREMENTS verifying the absence of the Nitrogen Backup System low pressure alarms. Tho, 2,1 hour. Froguency is bacod,, on, oninorn *udgmon in.÷ ,
| |
| .. , .* ,,o; of.. th fact,that. adoguato,.q i.+ndication of* ......... i" a-ailablo to th.
| |
| rt1 j-ns op.. rao
| |
| . and. t*hc Fro. ..uo c.. ha..also, boc . ho..... n to acccpt..... b o throuah oocratino exoorionco.
| |
| SR 3.6.1.5.2 Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is not present. This Surveillance is performed by observing local or control room indications of vacuum breaker position. Thc* 1, day, Fr...u..nc. is basc on.. ;n .... r
| |
| ~ jugmot, s cnsioro adcquat in ow of othcr indicaton ofivacuum, ll
| |
| ~~~jj-,' roac sau oailabo; tor op1r,,ion pors,;nno , and~ has*+; b.-n shown,,to Two Notes are added to this SR. The first Note allows reactor building-to-suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR.
| |
| These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. The second Note is included to clarify that vacuum breakers open due to an actual differential pressure are not considered as failing this SR.
| |
| SR 3.6.1.5.3 Each vacuum breaker must be cycled to ensure that it opens properly to perform its design function and returns to its fully closed position. This SR ensures that the safety analysis assumptions are valid. This is accomplished by manually verifying that each mechanical vacuum breaker is free to open and verifying each pneumatic butterfly valve operates through at least one complete cycle of full travel. Trhe -d-e-
| |
| [ Insert 1 Program .....i;..... to*'perfor
| |
| ...... t"oting* at least once cecry (continued)
| |
| Brunswick Unit 1 B 3.6.1.5-7
| |
| ...- eiinN.3 Revision No. 31 I
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.4 REQUIREMENTS (continued) Demonstration of vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of < 0.5 psid is valid. This is accomplished by demonstrating that the force required to open each mechanical vacuum breaker is < 0.5 psid and demonstrating that each pneumatic butterfly valve opens at _>0.4 psid and < 0.5 psid with drywell pressure negative with respect to reactor building pressure. Thc 21 month Frcquneny has be1 d emo, itratod to bc acceptable, bao onl oaIngI cxpricec and1 is fur. ... ,.*, ther jutfe ....
| |
| eas cfohrSreillance1s... pe,"formed mere SR 3.6.1.5.5 To ensure the pneumatic butterfly valves have sufficient capacity to actuate and cycle following a LOCA and subsequent primary containment isolation, Nitrogen Backup System leakage must be within the design limit.
| |
| This SR ensures that overall system leakage is within a design limit of 0.65 scfm. This is accomplished by measuring the nitrogen bottle supply pressure decrease while maintaining approximately 95 psig to the nitrogen backup subsystem during the test with an initial nitrogen bottle supply pressure of> 1130 psig. The, system leakag te is perfore ever,'....
| |
| I Insert 1 S the.Frequency.. is concude to.,be acceptab. from,.a..reliability standpoint.
| |
| SR 3.6.1.5.6 This SR ensures that in the event a LOCA and subsequent primary containment isolation occurs, the Nitrogen Backup System will actuate to perform its design function and supply nitrogen gas at the required pressure to the pneumatic operators of the butterfly valves. F-h~e
| |
| ('continued)
| |
| Brunswick Unit 1 B3.6.1.5-8 BrunwickUnit1 B.6.15-8Revision No. 36
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.6 (continued)
| |
| REQUIREMENTS oA 21,month Fro ........ is. baed, on,.the need, to ,-,oeorm this ure
| |
| ... lla'ne, I Insert ,,under, the con.t. n that;"appiy*"during a plant ou..tage. Operating REFERENCES 1.: NRC Generic Letter GL 84-09, Recombiner Capability Requirements of 10 CFR 50.44(c)(3)(ii).
| |
| : 2. UFSAR, Section 6.2.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.6.1.5-9 BrunwickUniNo. 31 I 1 B3.6.5-9Revision
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.6 BASES SURVEILLANCE SIR 3.6.1.6.1* (continued) /[In__sert 1 REQUIREMENTS and drywell is maintained > 0.5 it es the initial differential pressure for 1 hour without nitrogen makeup. T~he 11! day Frequency is based on engine...ering,,,
| |
| judgent
| |
| ... + i conidre.s,,, adequat in' vio'w of* o-hor indications be a...eptab*lo throu...h opertin exper...enc... This verification is also required within 6 hours after any discharge of steam to the suppression chamber from any source, and within 6 hours after an operation that causes any of the vacuum breakers to open.
| |
| A Note is added to this SR which allows suppression chamber-to-drywell I
| |
| vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers.
| |
| SR 3.6.1.6.2 Each required vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This is accomplished by verifying each required vacuum breaker operates through at least one complete cycle of full travel. This SR ensures that the safety analysis assumptions are valid. The-J2-day, Frequncyo-thi.S..wa developed, basod,-on Inse~st .. Testing
| |
| [LInsert Program requirements. to peror valve..... tetn tlatoc vr 82-davs.7 In addition, this functional test is required within 12 hours after a discharge of steam to the suppression chamber from the SRVs.
| |
| SR 3.6.1.6.3 Verification of the vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 0.5 psid is valid. The 21! month Frequency is I Insert 1 apply during a plant outage and the potential for an unplanned transient it the Sur:eillancc were performed with the reactor at power. The 21 month Frequency has been demonstrated to be acceptable, based on operating experience, and is further justified because of other sun'eillanees performod more frequently that convey the proper functioning status ot each vacuum breakar.
| |
| REFERENCES 1. UFSAR, Section 6.2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 ...-
| |
| B 3.6.1.6-5 eiinN.6 No. 65 Revision I
| |
| | |
| Suppression Pool Average Temperature B 3.6.2.1 BASES ACTIONS E.1 and E.2 (continued)
| |
| If suppression pool average temperature cannot be maintained at
| |
| _<120°F, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the reactor pressure must be reduced to
| |
| < 200 psig within 12 hours, and the plant must be brought to at least MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| Continued addition of heat to the suppression pool with suppression pool temperature > 120°F could result in exceeding the design basis maximum allowable values for primary containment temperature or pressure.
| |
| Furthermore, if a blowdown were to occur when the temperature was
| |
| > 120°F, the maximum allowable bulk and local temperatures could be exceeded very quickly.
| |
| SURVEILLANCE SIR 3.6.2.1.1 REQUIREMENTS The suppression pool average temperature is regularly monitored to ensure that the required limits are satisfied. The average temperature is determined using an algorithm with inputs from OPERABLE suppression pool water temperature channels. -Tho,2)1 hou Fr..oqu..nc.. has,been,,
| |
| IInsert 1 being ddod to the.upp.....on pool..by,,, tcs.ting, howo'-cr, it is nocossa,';
| |
| to monito
| |
| ;supp...... on ... pool p.....
| |
| .. ... mor frqenl... , The 5 minute Frequency during testing is justified by the rates at which tests will heat up the suppression pool, has been shown to be acceptable based on operating experience, and provides assurance that allowable pool temperatures are not exceeded. The F-e iueiefe-a=e further justified in view of other indications available in the co rol room, including alarms, to alert the operator to an abnormal suppressi on pool average temperature condiion.I.L_*Frequency is REFERENCES 1. NEDC-33039P, Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Extended Power Uprate, August 2001.
| |
| : 2. NUIREG-0783.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.6.2.1-5 No. 31 I BrunwickUnitI B.6.21-5Revision
| |
| | |
| Suppression Pool Water Level B 3.6.2.2 BASES ACTIONS B.1 and B.2 (continued)
| |
| If suppression pool water level cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS Verification of the suppression pool water level is to ensure that the required limits are satisfied. Thc. 21 hou Fr..qu...nc.. of this SR h. been
| |
| [ Insert 1the. 21 hour. Frequency is con.d.re a,4. cgu..4
| |
| ... in. view.of..ot.hcr indicat~ons a-vailable in the c..ntrol room, including "lams to. aler th-* e operator to a*n REFERENCES 1. UFSAR, Section 6.2.1.1.3.2.
| |
| : 2. NEDC-33039P, Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Extended Power Uprate, August 2001.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.6.2.2-3 BrunwickUnit1 B.6.22-3Revision No. 31
| |
| | |
| RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS (continued)
| |
| B.__1 With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours. In this condition, there is a substantial loss of the primary containment pressure and temperature mitigation function. The 8 hour Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to remove heat from primary containment are available.
| |
| C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply.
| |
| To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. 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 SInset fl-... inadvertently misaligned, such as check valves.
| |
| ..n... pr.....u....control, imprpcpr "aly
| |
| ...... o would....' affcct ol
| |
| ~ingc ~b~ytom th proabiityof.n....n ro q.u,,ring initiati..n.of....
| |
| (continued'*
| |
| Brunswick Unit 1 ...-
| |
| B 3.6.2.3-3 eiinN.4 No. 41 Revision I
| |
| | |
| RHR Suppression Pool Cooling B 3.6.2.3 BAS ES SURVEILLANCE SR 3.6.2.3.2 REQUIREMENTS (continued) Verifying that each RHR pump develops a flow rate _> 7700 gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that the primary containment pressure and temperature can be maintained below the design limits during a DBA (Ref. 2). The normal test of centrifugal pump performance required by ASME OM Code (Ref. 4) is covered by the requirements of LCO 3.5.1, "ECCS--Operating." This test confirms one point on the pump design curve, and the results are indicative of overall performance.
| |
| Such tests confirm component OPERABILITY, and detect incipient Thc ,Fre........
| |
| failures by indicating abnormal performance.__r',
| |
| * of **.
| |
| this SR i,s*,
| |
| oo .,d.*s, REFERENCES 1. UFSAR, Section 6.2.1.1.3.2.
| |
| : 2. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| : 4. ASME Code for Operation and Maintenance of Nuclear Power Plants.
| |
| Brunswick Unit 1I ...-
| |
| B 3.6.2.3-4 eiinN.5 No. 59 Revision I
| |
| | |
| Primary Containment Oxygen Concentration B 3.6.3.1 BASES APPLICABILITY 24 hours of a startup, or within the last 24 hours before a scheduled (continued) power reduction < 15% RTP, is low enough that these "windows," when the primary containment is not inerted, are also justified. The 24 hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting.
| |
| ACTIONS A.1 If oxygen concentration is Ž_4.0 v/o at any time while operating in MODE 1, with the exception of the relaxations allowed during startup and shutdown, oxygen concentration must be restored to < 4.0 v/o within 24 hours. The 24 hour Completion Time is allowed when oxygen concentration is > 4.0 v/o because of the availability of other hydrogen and oxygen mitigating systems and the low probability and long duration of an event that would generate significant amounts of hydrogen and oxygen occurring during this period.
| |
| B.__1 If oxygen concentration cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, power must be reduced to < 15% RTP within 8 hours. The 8 hour Completion Time is reasonable, based on operating experience, to reduce reactor power from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.6.3.1.1 REQUIREMENTS The primary containment must be determined to be inerted by verifying that oxygen concentration is < 4.0 v/o. -Tho....7,4*, da ,..........
| |
| .. od o.*,,n is bas....
| |
| thc, lo....ratc at which,. .......... concntrt.o c..n chango,*. .andon. othcr REFERENCES 1. UFSAR, Section 6.2.5.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.6.3.1-2 Brun wic Unt No. 66 I 1B36.31-2Revision
| |
| | |
| Secondary Containment B 3.6.4.1 BASES ACTIONS C.1 and C.2 (continued) assemblies would not be a sufficient reason to require a reactor shutdown.
| |
| SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and one secondary containment access door in each access opening are closed ensures that the infiltration of outside air of such magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. In this application, the term "sealed" has no connotation of leak tightness.
| |
| Maintaining secondary containment OPERABILITY requires verifying one door in each access opening is closed. The,2 mont ,,,h Fr......... for thcse SRs has bccn shown to bo adequate, bascd on op.......
| |
| exncricncc. rind is rcnnf1d~r~d adren'-ntr in* ,-inw nf nthnr indic~nt*nn~ of F i .................... "-I..................
| |
| %r SR 3.6.4.1.3 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will establish and maintain a negative pressure in the secondary containment. This is confirmed by demonstrating that one SGT subsystem can maintain >_0.25 inches of vacuum water gauge for 1 hour at a flow rate < 3000 cfm. The 1 hour test period allows secondary containment to be in thermal equilibrium at steady state conditions.
| |
| Therefore, this test is used to ensure secondary containment boundary integrity. Since this SR is a secondary containment test, it need not be performed with each SGT subsystem. Tho,- S-T sub,,1stem. + arc.. 'ltcstcd LInert tt...., .p......cxp.......ha.demo .trat .....s componen-ts will F*,,-lrequncy Thercfor, thoFlrequencywa conc~~ludod to* bc4 accetab (continued)
| |
| Brunswick Unit t B 3.6.4.1-4 BrunwickUnitINo. 31 I B.6.41-4Revision
| |
| | |
| SCIDs B 3.6.4.2 BASES ACTIONS C.1 and C.2 (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| 0.1 and 0.2 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
| |
| LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action 0.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
| |
| SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS Verifying that the isolation time of each automatic SCID is within limits, by cycling each SCID through one complete cycle of full travel and measuring the isolation time, is required to demonstrate OPERABILITY.
| |
| The isolation time test ensures that the SCID will isolate in the required time period. The. rq- nc of..... t*hi- SR i once. per.21 months.- Op..r.ating Sr.....lanc when..pe.... d at.t4.* he 21 month Frequenqcy. Therefore, thc.
| |
| Frequency ...... concluded to. be..... a.l from' a..reliability standpoin.t (continued' Brunswick Unit 1 B3.6.4.2-5 No. 31 I BrunwickUnitI B.6.42-5Revision
| |
| | |
| SCIDs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.2 REQUIREMENTS (continued) Verifying that each automatic SClD closes on a secondary containment isolation signal is required to minimize leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SClD will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCo 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function. Thc 21 month Frequcncy is based on the need to REFERENCES 1. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
| |
| : 2. UFSAR, Section 15.6.4.
| |
| : 3. Not used.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. Technical Requirements Manual.
| |
| Brunswick Unit 1 ...-
| |
| B3.6.4.2-6 eiinN.3 No. 31 Revision
| |
| | |
| SGT System B 3.6.4.3 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued) assemblies would not be a sufficient reason to require a reactor shutdown.
| |
| E.1 and E.2 When two SGT subsystems are inoperable, if applicable, movement of recently irradiated fuel assemblies in secondary containment must immediately, be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
| |
| Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
| |
| LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since recently irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
| |
| S URVE ILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem, by initiating (from the control room) flow through the HEPA filters and charcoal adsorbers, for _>10 continuous hours ensures that both subsystems 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 automatic control for
| |
| > 10 continuous hours evcry 3!-deys eliminates moisture on the adsrbenrsr andl HEIPA filters. Thc.- 31 day. Fr-cqu,,.ncy, ,wa. do,,,lopo in*A; (continued)
| |
| Brunswick Unit 1 ...-
| |
| B 3.6.4.3-5 eiinN.3 No. 31 I Revision
| |
| | |
| SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.2 REQUI REMENTS (continued) This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1 .52 (Ref. 6),
| |
| except as specified in Specification 5.5.7, "Ventilation Filter Testing Program (VFTP)". The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). It is noted that, per the basis provided by ESR 99-00055 (Ref. 7), system flow rate is determined using installed calibrated flow orifice plates. Specific test frequencies and additional information are discussed in detail in the VFTP.
| |
| SR 3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Su-,"cillancc can bc performed
| |
| [Insert 1 the.e.cmponent wil u..suall.."y pass the Sur--illance wh'en' performed at the., O1. mo*...*"nth ,Freunc'vn***.. The LOIC SYS'- VTEMI* 1IFUNCTIONIkAL TES*T in LC 3-,3.6-2,*"Secon~dar' Co"-t'inm... I solation Inqstrume~ntation,",
| |
| u':ur IiiU~. [iII~. ~ in I]E(]~2IflF~ rnrnninir ir~rinn O'T Tinc PaTeTIy TIuncion.
| |
| Thereofore, the Fequency..... wa found to bo a,* ;;;-;;:; * ;*;;*L;:;;*'v"
| |
| ;*L*L;UL-:;J;U st-anpi~t.
| |
| REFERENCES 1. UFSAR, Section 6.5.1.
| |
| : 2. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
| |
| : 3. UFSAR Section 15.6.4.
| |
| : 4. Not used.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. Regulatory Guide 1.52, Revision 1.
| |
| : 7. ESR 99-00055, SBGT and CBEAF Technical Specification Surveillance Flow Measurement.
| |
| Brunswick Unit 1 B 3.6.4.3-6 BrunwickUnit1 No. 31 I B.6.43-6Revision
| |
| | |
| RHRSW System B 3.7.1 BASES (continued)
| |
| SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual, power operated, and automatic valve in each RHRSW subsystem flow path provides assurance that the proper flow paths exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system.
| |
| 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
| |
| [ Inset 1
| |
| * inadvertently misaligned, such as check valves.
| |
| Thc 31 a F4r.quc,,c. is basod, on cngnccin ;,,dgm-cnt,* is .,-,i't;cnt REFERENCES 1. UFSAR, Section 9.2.1.2.
| |
| : 2. UFSAR, Chapter 6.2.
| |
| : 3. N EDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995; and Supplement 1, March 1996.
| |
| : 4. Letter BR5-96-074, Long Term Suppression Pool Temperature-Suppression Pool Cooling Mode for Long Term Containment Cooling, from M. E. Ball (GE) to R. E. Helme (CP&L),
| |
| September 19, 1996.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. GENE-B2100565-09, Technical Specification Improvements to the Emergency Core Cooling System for the Carolina Power and Light Brunswick Steam Electric Plant Units 1 and 2, Revision 1, October 1996.
| |
| Brunswick Unit 1 ..-
| |
| B 3.7.1-5 eiinN.4 No. 41 Revision I
| |
| | |
| SW System and UHS B 3.7.2 BASES ACTIONS 1.1 and 1.2 (continued)
| |
| If Required Actions cannot be completed within the associated Completion Time of Condition A, B, D, E, F, G, and H; Required Action C.2 cannot be completed within the associated Completion Time; two or more required NSW pumps are inoperable; the SW System is inoperable for reasons other than Conditions A, B, C, D, E, F, and G; or the UHS is inoperable for reasons other than Condition H (e.g., low water level); 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 SURVEILANCE within 12 hours and in MODE 4 within 36 hours. 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.
| |
| SR 3.7.2.1 REQUIREMENTS This SR verifies the water level in the SW pump suction bay of the intake structure to be sufficient for the proper operation of the SW pumps (net positive suction head and pump vortexing are considered in determining this limit). This SR may be accomplished by measuring intake canal water level provided the deviation in water level between the intake canal and the pump suction bay due to the differential pressure of the traveling L nsrt1 ,* screens is taken into account. Tho, 21, hou Fr..equ....c. i's basod, on durng-hcb;'lk. . Ippicable.-, ODES..,,AII1 SR 3.7.2.2 Verification of the UHS temperature ensures that the heat removal capability of the SW System is within the assumptions of the DBA analysis.. Thc, 21t hour Frqcn i..... bas- on*,oper.tin ep...rience,.,
| |
| (continued')
| |
| Brunswick Unit 1 ..- 1ReiinN.3 B 3.7.2-11 Revision No. 31 I
| |
| | |
| SW System and UHS B 3.7.2 BASES SURVEILLANCE SR 3.7.2.3 REQUI REME NTS (continued) Verifying the correct alignment for each manual, power operated, and automatic valve in the SW System flow paths provide assurance that the proper flow paths will exist for SW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time.
| |
| This SR does not require 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 inadvertently misaligned, such as check valves.
| |
| This SR is modified by a Note indicating that isolation of the SW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the SW System. As such, when all SW pumps, valves, and piping are OPERABLE, but a branch connection off the NSW or CSW header is isolated, the SW LLnsrK1J*. System is still OPERABLE.
| |
| withtho rocd conrol
| |
| .... govcrning ..al..oprati..n...nd..n..ure.
| |
| SR 3.7.2,4 The dominant contributor to a loss of DG cooling is a failure of the normal and alternate cooling water supply valves to open on demand from their normally closed position. As a result, since only three site NSW pumps are required to be OPERABLE, the capability to automatically transfer the cooling water supply to the OG jacket water coolers from the NSW header of one unit to the NSW header of the opposite unit is necessary to meet single failure criteria.
| |
| (continued' Brunswick Unit 1 B3721 B 3.7.2-12 eiinN.3 No. 31 I Revision
| |
| | |
| SW System and UHS B 3.7.2 BASES SR 3.7.2.4 (continued)
| |
| REQUIREMENTS
| |
| [Insert 1I ln..... :e Toting PDrogram Fr-equenc... for testring of ;'alres.
| |
| To minimize testing of the cooling water supply valves to each DG, Note 1 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. Note 2 indicates that isolation of the SW System to a OG renders the DG inoperable but does not affect the OPERABILITY of the SW System. As such, ifthe automatic transfer of the cooling water supply valves associated with a DG fails this Surveillance, the DG should be considered inoperable. However, the SW System is still OPERABLE.
| |
| It is not necessary to declare the DG inoperable if the service water supply valves to the affected DG are administratively controlled to ensure cooling water is supplied to the DG and two NSW pumps are operable on the corresponding NSW header that the DG is aligned to. This ensures that a single active failure will not result in more than one DG not receiving cooling water (Ref. 5).
| |
| SR 3.7.2.5 This SR verifies that the automatic isolation valves of the SW System will automatically align to the safety or emergency position to provide cooling water exclusively to the safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal. This SR also verifies the automatic start capability of the required NSW pumps.
| |
| ]Insert 1 usulen'lynr,**
| |
| pass.tel SR whe-n prfo.**rme at* the 2"1 monthfl F~requency.~
| |
| (continued)
| |
| Brunswick Unit 1 ..-
| |
| B 3.7.2-13 3ReiinN.3 Revision No. 31 I
| |
| | |
| CREV System B 3.7.3 BASES (continued)
| |
| ACTIONS E.1, E.2, and E.3 (continued)
| |
| The Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, 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.
| |
| During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, with two CREV subsystems inoperable or with one or more CREV subsystems inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the CRE. These actions place the unit in a condition that minimizes the accident risk.
| |
| If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
| |
| SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies that each CREV subsystem in a standby mode starts on demand and continues to operate. This SR includes initiating flowI through the associated HEPA filter and charcoal adsorber bank. Standby systems should be checked periodically to ensure that they start and function properly. A~s the en'-ironmental and normal oporating conditions of* this.... o. ar not.. e....., testing eac,,h subsystem once...e..
| |
| 31 day pro..id.. an. adequate check.......
| |
| on. thi".... tem. Since the CREV subsystems do not have installed heaters, each subsystem need only be nsrt 1operaeatedf orŽ15minutesto stodemonstrate thefunction ofthe subsystem.
| |
| The,314 day,Fr..quency. is basd,, on the kno...n reliability-'
| |
| of"the cqu....t (continued)
| |
| Brunswick Unit 1 B3.7.3-6 Brunwic Uni No. 64 I I 3.73-6Revision
| |
| | |
| CREV System B 3.7.3 BASES SURVEILLANCE SR 3.7.3.4 REQUIREMENTS (continued) This SR verifies that on an actual or simulated initiation signal, each CREV subsystem starts and operates. This SR includes ensuring outside air flow is diverted to the HEPA filter and charcoal adsorber bank of each CREV subsystem. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.7.1 overlaps this SR to provide complete testing of the safety inerl-,,. .,,,,,function. Opor.t.n c."pc....cn... has, dcm....trated, that thc ..o.....nts REFERENCES 1. UFSAR, Section 6.4.
| |
| : 2. UFSAR, Section 9.4.
| |
| : 3. UFSAR, Section 6.4.4.1.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| : 5. ESR 99-00055, SBGT and CBEAF Technical Specification Surveillance Flow Measurement.
| |
| : 6. Regulatory Guide 1.196
| |
| : 7. NEI 99-03, "Control Room Habitability Assessment," June 2001.
| |
| : 8. Letter from Eric J. Leeds (NRC) to James W. Davis (N El) 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).
| |
| Brunswick Unit 1 B 3.7.3-8 Brunwic Uni No. 64 I 1 3.73-8Revision
| |
| | |
| Control Room AC System B 3.7.4 BASES ACTIONS F.1, F.2. and F.3 (continued)
| |
| LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition F are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuelf movement is independent of reactor operations.
| |
| Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.
| |
| During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, with three control room AC subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.
| |
| If applicable, CORE ALTERATIONS and handling of irradiated fuel in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until the OPDRVs are suspended.
| |
| SURVEILLANCE SR 3.7.4.1 REQUI REM ENTS This SR verifies that the heat removal capability of the system is sufficient to remove the control room heat load assumed in the safety analyses.
| |
| ] Inset 1
| |
| * The SR consists of a combination of testing and calculation. The 2month,,,k*Frqoc
| |
| ..... .... -,t -.',,-, z-ig-,,f,--nt dcgradation of th,,
| |
| "ppop REFERENCES 1. UFSAR, Section 6.4.2.
| |
| : 2. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.7.4-5 Brunwic Uni No. 31 I I 3.74-5Revision
| |
| | |
| Main Condenser Offgas B 3.7.5 BASES ACTIONS B.1, B.2, B.3.1, and B.3.2 (continued)
| |
| An alternative to Required Actions B.1 and B.2 is to place the unit 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 12 hours and in MODE 4 within 36 hours. 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 3.7.5.1 REQUIREMENTS This SR, on,,*a'*31'""
| |
| da Frqec,,..
| |
| , requires an isotopic analysis of an offgas sample (taken at the discharge of the main condenser air ejector prior to dilution or discharge) to ensure that the required limits are satisfied. The noble gases to be sampled are Xe-i133, Xe-i135, Xe-i138, Kr-85m, Kr-87, and Kr-88. If the measured rate of radioactivity increases significantly (by _ 50% after correcting for expected increases due to changes in THERMAL POWER), an isotopic analysis is also performed within 4 hours after the increase is indicated (by the condenser air ejector noble gas activity monitor), to ensure that the increase is not indicative of I nsr 1
| |
| * a sustained increase in the radioactivity rate. Thc 3 d ay, FounyIs
| |
| ~'n vcw acquto f ohcrinstumctaton h.tconinuu.l moitrcquecy, This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after any main steam line is not isolated and the SJAE is in operation. Only in this condition can radioactive fission gases be in the Main Condenser Offgas System at significant rates.
| |
| REFERENCES 1. UFSAR, Section 11.3.
| |
| : 2. 10 CFR 50.67.
| |
| : 3. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.7.5-3 Brunwic Uni No. 31 I I 3.75-3Revision
| |
| | |
| Main Turbine Bypass System B 3.7.6 BASES ACTIONS B.__
| |
| (continued)
| |
| If the Main Turbine Bypass System cannot be restored to OPERABLE status and the APLHGR, MCPR, and LHGR limits for an inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to < 23% RTP. As discussed in the Applicability section, operation at < 23% RTP results in sufficient margin to the required limits, and the Main Turbine Bypass System is not required to protect fuel integrity during the applicable safety analyses transients. The 4 hour Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
| |
| SR 3.7.6.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The 31* day Frcquonc..i. ba"c"o", n
| |
| .....ufac"tu.... ' recommondations (Rcf.f 6) is c.. oniten ,,ith th,,
| |
| proceduralw,. contro-"s go....erning,,;'- valve operation, and"-, ensures correct '"al-ve,.
| |
| u.ually pas the..SR.when performed at the, 31 da-y, Frequency.
| |
| [Inerll-*
| |
| SR 3.7.6.2 The Main Turbine Bypass System is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system initiation signals, the valves will actuate to their required position.
| |
| The 0/I mont,,,-h Frequen,,.cy.., is. based, on the need*,*, to,- pe,,for*,m,- this Su,"veillaneo une t*4' he.. conditins-that applyh du,rin a...uitutg a"nd4 because o,.,f the,, potentiafl "foran un'.nlln od.'.,-.,,
| |
| transieint÷ if: the S.,urv llane.,.',
| |
| ,were performed wi-th the reactor at power.
| |
| (continued')
| |
| ,i .. ... . ... r Brunswick Unit 1 B 3.7.6-3 Brunwic Uni No. 58 I 1 3.76-3Revision
| |
| | |
| Main Turbine Bypass System B 3.7.6 BASES SURVEILLANCE SR 3.7.6.3 REQUIREMENTS
| |
| ('continued) This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety
| |
| [ Insert 1 analysis. The response time limits are specified in a design calculation.
| |
| SThc 21! month Frequ-encv is based on thc need to per"fo rm this Surv.ll.nc under...t,,he coniton thf,.
| |
| at* apps
| |
| ,*l, d- a ui-*t otg
| |
| *urinl n becus of...the potenti.al for an unplanned-.. tran.ien if*the Sureill..ance...
| |
| REFERENCES 1. UFSAR, Section 7.7.1.4.
| |
| : 2. UFSAR, Section 15.2.1.
| |
| : 3. UFSAR, Section 15.2.2.
| |
| : 4. UFSAR, Section 15.1.2.
| |
| : 5. 10 CFR 50.36(c)(2)(ii).
| |
| : 6. GE Service Information Letter No. 413, Main Steam Bypass Valve Testing, October 4, 1984.
| |
| Brunswick Unit 1 B3.7.6-4 Brunwic Uni No. 86 I I 3.76-4Revision
| |
| | |
| Spend Fuel Storage Pool Water Level B 3.7.7 BASES (continued)
| |
| LCO The specified water level (19 feet 11 inches above the top of irradiated fuel assemblies seated in the High Density Fuel Storage System racks) preserves the assumptions of the fuel handling accident analysis (Ref. 2).
| |
| As such, the water level is the minimum required for fuel movement within the spent fuel storage pool. This water level corresponds to 36 feet 11 3/4 inches spent fuel storage pool level.
| |
| APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool since the potential for a release of fission products exists.
| |
| ACTIONS A.11 LCO 3.0.3 is not applicable while in MODE 4 or 5. However, since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown.
| |
| When the initial conditions for an accident cannot be met, action must be taken to preclude the accident from occurring. Ifthe spent fuel storage pool level is less than required, the movement of irradiated fuel assemblies in the spent fuel storage pool is suspended immediately.
| |
| Suspension of this activity shall not preclude completion of movement of an irradiated fuel assembly to a safe position. This effectively precludes a spent fuel handling accident from occurring.
| |
| SURVEILLANCE SR 3.7.7.1 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in the spent fuel storage pool must be Insert 1 checked periodically. Tho., 7.d.y,Frcq-cncy ic .... abl .....
| |
| *,..** c,.n.idcring spent fuc'" stoag pol....nictr.adasoitd lr (continued)
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| Brunswick Unit 1 ..-
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| B 3.7.7-2 eiinN.3 No. 31 [
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| Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE recommendations of Safety Guide 9 (Ref. 5), Regulatory Guide 1.9 REQUIREMENTS (Ref. 11), and Regulatory Guide 1.137 (Ref. 12), as addressed in the (continued) UFSAR.
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| Where the SIRs discussed herein specify voltage and frequency tolerances, the following summary is applicable. The minimum steady state output voltage of 3750 V is derived from the recommendations found in Safety Guide 9 (Ref. 5) and bounds the minimum steady state output voltage criteria of 3621 V associated with the 4.16 kV. emergency buses analyzed in the AC Auxiliary Electrical Distribution System Study.
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| This value (3621 V) allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 3600 V. It also allows for voltage drops to motors and other equipment down through the 480 V level where minimum operating voltage is also usually specified as 90% of name plate rating. The specified maximum steady state output voltage of 4300 V ensures the maximum operating voltage at the safety related 480 V substations is no more than the maximum rated steady state voltage criteria for the 480 V motor control centers. The maximum steady state output voltage was .determined taking into consideration the voltage drop between the DGs and the 4.16 kV emergency buses and a 5% voltage boost at the 480 V substation transformers. This maximum steady state output voltage also ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated steady state operating voltage. The specified minimum and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to + 2% of the 60 Hz nominal frequency and are derived from the recommendations found in Safety Guide 9 (Ref. 5).
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| 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 that appropriate
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| [ Insert 1
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| * independence of offsite circuits is maintained. The, 7,-,da F......... i' (continued)
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| Brunswick Unit 1 B 3.8.1-20 BrunwickUniNo. 31 I I B3.8.-20Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 REQU IREM ENTS (continued) These SRs help to ensure the availability of the standby electrical power supply to mitigate OBAs and transients and maintain the unit in a safe shutdown condition.
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| To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs have been modified by a Note (Note 1 for SR 3.8.1.2 and SR 3.8.1.7) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period.
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| For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained.
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| In order to reduce stress and wear on diesel engines, some manufacturers recommend 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 2 of SR 3.8.1.2.
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| SR 3.8.1.7 requires that, at "a181 da., Fr........, the OG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The minimum voltage and frequency stated in the SR are those necessary to ensure the OG can accept DBA loading while maintaining acceptable voltage and frequency levels. Stable operation at the nominal voltage and frequency values is also essential to establishing DG OPERABILITY, but a time constraint is not imposed. This is because a typical DG will experience a period of voltage and frequency oscillations prior to reaching steady state operation if these oscillations are not dampened by load application. This period may be extended beyond the 10 second acceptance criteria and could be cause for failing the SR. In lieu of a time constraint in the SR, BNP will monitor and trend the actual time to reach steady state operation as a means of ensuring there is no voltage regulator or governor degradation which could cause a OG to become inoperable. The 10 second start requirement supports and is conservative with respect to the assumptions in the design basis LOCA analysis of UPSAR, Section 6.3 (Ref. 6). The 10 second start requirement is not applicable to SR 3.8.1.2 (see Note 2 of SR 3.8.1.2),
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| (continued)
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| Brunswick Unit 1 B 3.8.1-21 BrunwickUniNo. 31 I 1 B3.8.-21Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (continued)
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| REQUIREMENTS when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.7 applies.
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| To minimize testing of the DGs, Note 3 to SR 3.8.1.2 and Note 2 to SR 3.8.1.7 allow a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on'either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit.
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| The* 3491..d,,y Fr,.qucn,-y for,SR "381. i cons,-,-istent* wit:h R,,,-u.I,,÷,r-'
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| Gui_.9; (RcIDf. 11) Thc* 181 day Frequency for SR 3.81. is a reduct..*...ion in, cold,, t,,s..',g consistent with Gcnoric Letter"81! 15* (Ref. 10). These*.
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| SR 3.8.1.3 This Surveillance verifies that the DGs are capable of synchronizing and accepting a load approximately equivalent to the continuous rating of the DGs. 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.
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| Although DG is normally no power factor atrequirements operated arebetween a power factor established 0.8 by this SR, lagging and the 1.0.
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| The 0.8 value is the design rating of the machine, while 1.0 is the generator design limitation which if exceeded could lead to generator instability while in parallel with the offsite circuit. The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in order to maintain DG
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| [ Inset 1
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| * OPERABILITY.
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| (continued)
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| Brunswick Unit 1 ..-
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| B 3.8.1-22 2ReiinN.3 Revision No. 31 J
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 (continued)
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| REQUIREMENTS Note 1 modifies this Surveillance to indicate that diesel engine runs for this Surveillance may include gradual loading so that mechanical stress and wear on the diesel engine are minimized.
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| Note 2 modifies this Surveillance by stating that momentary transients because of changing bus loads do not invalidate this test. Similarly, momentary power factor transients outside the range normally used during the performance of this Surveillance do not invalidate the test.
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| 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.
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| Note 4 stipulates a prerequisite requirement for performance of this SR.
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| A successful DG start must precede this test to credit satisfactory performance.
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| To minimize testing of the DGs, Note 5 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit.
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| SR 3.8.1.4 This SR provides verification that the level of fuel oil in the engine mounted tank is slightly below the level at which the backup fuel oil transfer pump automatically starts. The level is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel oil for approximately 30 minutes of DG operation at rated load. This SR may I Inset 1
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| * be satisfied by verifying the absence of the associated low level alarm.
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| fu...oil.i a.. .....bo, c..lo
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| .... ........alarm arc* providod and facilt (continued'l Brunswick Unit 1 B3812 B 3.8.1-23 eiinN.3 No. 31 I Revision
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| | |
| AC Sources--Operating B 3:8.1 BASES SURVEILLANCE REQUIREMENTS SR 3.8.1.5 /*Periodic removal]
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| (continued) Microbiological fouling is a major cause of fuel o~i degradation. There are numerous bacteria that can grow in fuel oil and *use fouling, but all must have a water environment in order to survive. Rcmc'-al of water from the engine mounted tanks once..avery 31*"day... 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, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent
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| ',.hor, ~ I r1 rcr
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| *,'inI',JVr Il ~ ,'UrcrLv Il V .r VA
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| ^,f~ L I,"
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| III lii~i IL.; I I,*
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| 'z .JU, mlAn L Inser 1
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| * provides data regarding the watertight integrity of the fuel oil system. T-he
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| *.**.) This SR is for preventive maintenance. The presence of water (RfD42..*
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| does not necessarily represent a failure of this SR provided that accumulated water is removed during performance of this Surveillance.
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| Removal of accumulated water may be accomplished by draining a portion of fuel oil from the engine mounted fuel oil tank to the day fuel oil storage tank and draining any accumulated water from the day fuel oil storage tank in accordance with SR 3.8.3.3. The draining evolution will continue until accumulated water is verified to be removed from the engine mounted fuel oil tank.
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| SR 3.8.1.6 This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated storage tank to its associated day tank. It is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for fuel transfer systems are OPERABLE.
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| (continued)
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| Brunswick Unit 1 B 3.8.1-24 BrunwickUni No. 31 I I B3.8.-24Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 (continued)
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| REQUIREMENTS Thc, F-r..u..nc. for this
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| * SR i consisten
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| .. ,ith* the Frcq ucncy for te.sting the,
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| [Insert 1 nO - in SR 3"8.1.3 DG operation;* for SR 3.8.1. is;*normally long"cnough th.at fuc.. oil loe in.the
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| * engino mounted.. tank will be reduce to. the po. nt wAhere.' the fue ,l t-1 rnnsfnr nimn'n', ,+ijt'om'tim-lh, et-nrt' t,.' rc*'-iore fu el oill
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| .............. i ......................
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| Icycl in the cnginc mou-nted tank.
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| SR 3.8.1.8 Transfer of each 4.16 kV emergency bus power supply from the normal circuit to the preferred offsite circuit and from the preferred offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the offsite circuit distribution network to power the shutdown loads. In lieu of actually initiating an automatic circuit transfer, testing that adequately shows the capability of the transfer is acceptable. The automatic transfer testing may include any series of sequential, overlapping, or total steps so that the entire transfer sequence is verified. Thc 21 month Frequency of l Insert 1 c..nsideration, the. plant. conditions required to perform the* Sur'-'cillanco, end isinene to,...., be. c...ns.istont.with expected fu' c'cl lengths.'+*'
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| This SR is modified by three Notes. The reason for Note 1 is that, during operation with the reactor critical, performance of SR 3.8.1 .8.a, verification of automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit, could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems. Note 1 is not applicable to SR 3.8.1 .8.b, verification of manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsite circuit, since this evolution does not cause perturbations of the electrical distribution systems. Due to the shared configuration of certain systems (required to mitigate DBAs and transients) between BNP Units 1 and 2, both units' offsite circuits are required to be OPERABLE to supply power (continued'l Brunswick Unit 1 B 3.8.1-25 No. 31 I BrunwickUni 1 B3.8.-25Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued)
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| REQUIREMENTS
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| : b. Tripping its associated core spray pump with the DG solely supplying the bus.
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| The load rejection test is acceptable if the increase in diesel speed does Insert 1
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| * not exceed the overspeed trip setpoint. Th,, 2)1 month,* F-r........ .',
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| This SR is modified by three Notes. The reason for Note 1 is that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems.
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| Due to the shared configuration of certain systems (required to mitigate DBAs and transients) between BNP Units 1 and 2, all four DGs are required to be OPERABLE to supply power to these systems when either one or both units are in MODE 1, 2, or 3. In order to reduce the consequences of a potential perturbation to the electrical distribution systems during the performance of this Surveillance, while at the same time avoiding the need to shutdown both units to perform this Surveillance, Note 1 only precludes satisfying this Surveillance Requirement for DG 1 and DG 2 when Unit 1 is in MODE 1, 2, or 3.
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| During the performance of this Surveillance with Unit 1 not in MODE 1, 2, or 3 and with Unit 2 in MODE 1, 2, or 3; the applicable ACTIONS of the Unit 1 and Unit 2 Technical Specifications must be entered if DG 1 or DG 2 is rendered inoperable by the performance of this Surveillance.
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| Credit may be taken for unplanned events that satisfy this SR. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, Note 2 requires that, if synchronized to offsite power, testing must be performed using a power factor *( 0.9.
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| This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience. To minimize testing of the DGs, Note 3 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit.
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| (continued'*
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| Brunswick Unit 1 B 3.8.1-27 BrunwickUniNo. 31 I I B3.8.-27Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.10 REQUIREMENTS (continued) Consistent with Regulatory Guide 1.9 (Ref. 11), paragraph C.2.2.12, this Surveillance demonstrates that DG non-critical protective functions (e.g.,
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| high jacket water temperature) are bypassed on an ECCS initiation test signal and critical protective functions (engine overspeed, generator differential overcurrent, low lubricating oil pressure, reverse power, loss of field, and phase overcurrent-voltage restrained) trip the DG to avert substantial damage to the DG unit. The non-critical 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 OBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.
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| I Insert 1 Thc, '1 mont,,h Frcquoncy ic bazcd o.n ,enginc..ering ,,ugmc,,, takc* into, w~hn p...o. .. o t t hc 24 month Fre.o.c...*' Thercforc, thc Fr...uenc..
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| The SR is modified by a Note. To minimize testing of the D~s, the Note allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit.
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| SR 3.8.1.11 Brunswick Nuclear Plant performs a 60 minute run greater than or equal to the continuous rating (3500 kW) and less than or equal to the 2000-hour rating (3850 kW) to demonstrate diesel generator operation and to detect potential degradations and incipient failures. Post-accident DC loading is allowed up to the 2000-hour rating; however, it is impractical to require testing at this load but not to exceed it. A load band is provided to avoid routine overloading of the DC. Routine overloading may result in more frequent teardown inspections in order to maintain DG OPERABILITY. The DC starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelube and warmup, discussed in the Bases for SR 3.8.1.2, and for gradual loading, discussed in the Bases for SR 3.8.1 .3, are applicable to this SR.
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| (continued)
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| Brunswick Unit 1 B 3.8.1-28 Bruswik No. 48 I Uit B.8.-28Revision
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.11 (continued)
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| REQUIREMENTS 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 < 0.9. This power factor is chosen to be representative of
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| [ Inset 1
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| * the actual design basis inductive loading that the DG could experience.
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| The 21 month Frcqu-cncy is consistent 'with the recommondations of Rcg... tor÷,' Guid 1.4.9 (Rcf. 11), Table 1; takcs into consideration plant+
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| r~~c+nsitnt-wit+h cxprc+tcd4 ful cycrr~trlo lcngths.~h This Surveillance has been modified by two Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test. To minimize testing of the DGs, Note 2 allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit.
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| SR 3.8.1.12 Consistent with Regulatory Guide 1.9 (Ref. 11 ), paragraph C.2.2.13, demonstration of the test mode override feature ensures that the DG availability under accident conditions is not compromised as the result of testing. Interlocks to the LOCA sensing circuits cause the DG to automatically reset to ready-to-load operation if an ECCS initiation (continued)
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| Brunswick Unit 1 B 3.8.1-29 Bruswik I Uit B.8.-29Revision No. 48
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)
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| REQUIREMENTS signal is received during operation in the test mode. Ready-to-load 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 IEEE-308 (Ref. 13), paragraph 6.2.4(6).
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| In lieu of actually returning the DG to ready-to-load status, testing that adequately shows the capability of the DG to perform this function is acceptable. This testing may include any series of sequential, Iner 1overlapping, or total steps so that the entire sequence is verified.
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| This SR is modified by a Note. To minimize testing of the DGs, the Note allows a single test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit. If the DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit.
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| SR 3.8.1.13 Under accident conditions loads are sequentially connected to the bus by the automatic load sequence time delay relays. 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 10%
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| load sequence time interval tolerance 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. Reference 4 provides a summary of the automatic loading of ESF buses.
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| (.continued).
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| Brunswick Unit 1 B 3.8.1-30 BrunwickUniNo. 31 I 1 B3.8.-30Revision
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| | |
| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.13 (continued)
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| REQUIREMENTS Iner 1Rogulator*"'* Gud 1. (,Rof. 11.), Trblc .1; tak,,s intocownsidrto pl...nt consi-st-ont waith xpected*,- fuel cycle* lengths.
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| 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.
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| Due to the shared configuration of certain systems (required to mitigate DBAs and transients) between 1BNP Units 1 and 2, all four DGs, and associated load sequence relays, are required to be OPERABLE to supply power to these systems when either one or both units are in MODE 1, 2, or 3. In order to reduce potential consequences associated with removing a required offsite circuit from service during the performance of this Surveillance, reduce consequences of a potential perturbation to the electrical distribution systems during the performance of this Surveillance, and reduce challenges to safety systems, while at the same time avoiding the need to shutdown both units to perform this Surveillance, the Note only precludes satisfying this Surveillance Requirement for the load sequence relays associated with DG 1 and DG 2 when Unit 1 is in MODE 1, 2, or 3. During the performance of this Surveillance with Unit 1 not in MODE 1, 2, or 3 and with Unit 2 in MODE 1, 2, or 3; the applicable ACTIONS of the Unit 1 and Unit 2 Technical Specifications must be entered if a required offsite circuit, DG 1, or DG 2 is rendered inoperable by the performance of this Surveillance. Credit may be taken for unplanned events that satisfy this SR.
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| SR 3.8.1.14 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, ROS, and containment design limits are not exceeded.
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| This Surveillance demonstrates DG operation during a loss of offsite power actuation test signal in conjunction with an ECCS initiation signal.
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| This test verifies all actions encountered from the event, including shedding of the nonessential loads and energization of the emergency (continued'/
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| Brunswick Unit 1 B 3.8.1-31 BrunwickUni I B3.8.-31Revision No. 31
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| AC Sources--Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.14 (continued)
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| REQUIREMENTS 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.
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| The 10.5 second time period, which is allowed for the DG to auto-start and connect to its respective emergency bus, is conservatively derived from requirements of the accident analysis for responding 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 has been achieved.
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| The requirement to verify the connection and power supply of permanent and auto-connected 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 systems are not capable of being operated at full flow, or RHR systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
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| 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.
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| I Insert 1 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 being continuously circulated and temperature maintained consistent with procedural guidance. 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 (continued)
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| Brunswick Unit 1 B3.8.1-32 No. 31 I BrunwickUni I B3.8.-32Revision
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| Diesel Fuel Oil B 3.8.3 BASES ACTIONS D.1 (continued)
| |
| With a Required Action and associated Completion Time of Condition A, B, or C not met, or the stored diesel fuel oil not within limits for reasons other than addressed by Conditions A, B, or C, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.
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| SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for approximately 7 days at rated load. The approximate 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. For the purposes of this SR, the verification of the main fuel oil storage tank fuel oil volume is performed on a per DG basis. This per DG volume is obtained using the following equation:
| |
| -vL N- Uvo
| |
| ;where MvoL = measured fuel oil volume of the main fuel oil storage tank, UvoL = unusable fuel oil volume of the main fuel oil storage tank, and NDG --- number of DGs required to be OPERABLE.
| |
| The results from this equation must be _>20,850 gallons in order to satisfy
| |
| ] Inert*.*the acceptance criteria of SR 3.8.3.1.b.
| |
| .. ould be4aa..... of ..ny lg u...... of fuel oi du..ring this pcriod.
| |
| (continued)
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| Brunswick Unit 1 ..-
| |
| B 3.8.3-5 eiinN.5 No. 51 Revision I
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| | |
| Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE SR 3.8.3.2 (continued)
| |
| REQUIREMENTS accordance with ASTM D975-06b (Ref. 7). The 31 day period is acceptable because the fuel oil properties of interest, even ifthey were not within limits, would not have an immediate effect on DG operations.
| |
| This Surveillance ensures the availability of high quality fuel oil for the DGs.
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| SR 38.3. *
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| * Periodic removalI Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and 6ause fouling, but all must have a water environment in order to survive. RemevaI of water from the fuel storage tanks oncc over'; 31 dayc 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 OG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of Snsrlfl-~-*. the fuel oil system. The SuR.....nc F"r... qu....... i,. cs,,blch,, by, Rogulato,-y Guide 1.137 (Rcf. 2). This SR is for preventive maintenance.
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| The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.
| |
| REFERENCES 1. UFSAR, Section 8.3.1.1.6.2.8.
| |
| : 2. Regulatory Guide 1.137, January 1978.
| |
| : 3. UFSAR, Section 1.8.
| |
| : 4. UFSAR, Chapter 6.
| |
| : 5. UFSAR, Chapter 15.
| |
| : 6. 10 CFR 50.36(c)(2)(ii).
| |
| : 7. ASTM Standards: D4057-06; D975-06b; and D6217-98(2003)e1.
| |
| Brunswick Unit 1 B 3.8.3-7 Brunwic Uni I I 3.83-7Revision No. 75
| |
| | |
| DC Sources--Operating B 3.8.4 BASES ACTIONS B.1 and B.2 (continued) operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The Completion Time to bring the unit to MODE 4 is consistent with the time required in Regulatory Guide 1.93 (Ref. 7).
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| SURVEILLANCE SR 3.8.4.1 REQUI REM ENTS 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 and maintain the battery in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery. Thc 7 day Frequcncy is
| |
| - - - -... -A.'-
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| curnrnnrnju':n WICInn cnmnzircu wi~nn m~~un!iTcTIrcr reuunmmnnnriuuun Inn[n IEEE 15 RcD-f. 8)
| |
| SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell and inter-rack connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.
| |
| The connection resistance limits are < 1.2 times the established benchmark resistance values for the connections or < 5pohms above the established benchmark resistance values for the connections, whichever is higher. These connection resistance acceptance criteria were derived from IEEE-450 (Ref. 8) and IEEE-484 (Ref. 9), respectively.
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| I Insert1 *,
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| Thc Frequency for these can cause power loscos Frequency ic consistent 'with ..... acturor+ recommendations.'""'
| |
| (continued)
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| Brunswick Unit 1 B 3.8.4-5 Brunwic Uni No. 31 I 1 3.84-5Revision
| |
| | |
| DC Sources--Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.3 REQUIREMENTS (continued) 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).
| |
| FInset1Wi u5uIII n lllvnn~ II thr II*" ISR ncrfrm ih* nrVt III I R*U nnth FmI I ellr I' I j * :*,,v *,,,
| |
| *-, ,,::* v-* *,, ,, -,, ,,.. , . ,-. . ,. . * ,,1 * ,,* ,,J*
| |
| Th...fore, the......q..... rwa. ........... to.b...cc..t.b.o from.. a.
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| SR 3.8.4.4 Visual inspection of inter-cell and inter-rack connections provides an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The anti-corrosion material 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.
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| 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 this Surveillance.
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| -Ine rt 1 - Thea j*d .......cn Oprtn .... experience has sho...n that these components usually pass the SR when nur~uu,......... perfo u~ un rmediu Iy atiuthe 18}p. month uJ u Frequenc....
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| a-c... .. i--,.Iu tu...
| |
| I
| |
| * A HA*APA
| |
| * A I ..........
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| reliability ta ;ndpit SR 3.8.4.5 Battery charger capability requirements are derived from the design capacity of the chargers. According to Reference 3, the battery charger supply is required to be based on the largest combined demands of the (continued)
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| Brunswick Unit 1 B 3.8.4-6 BrunwicUni No. 31 I 1 3.84-6Revision
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| | |
| DC Sources--Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.5 (continued)
| |
| REQUIREMENTS various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, under any load condition. The minimum required amperes and duration ensures that these requirements can be satisfied.
| |
| LInert Th-,, f,.q~vee.} is acceptable, givcn battcr; charger
| |
| * re-;-l-ablity, ,-ndr tho I
| |
| U*.'U;,;;I* ;.U L;I;L.,.':;L;
| |
| * per,1"1r.....c is intende to- irt be ng..
| |
| consistcnt rontt'*÷ intervls.*
| |
| t'hi,"s. ".1Iwith expvvcted s h-Ina,-dittio ,tfltl. s 1hqe fuel" cyclo lengths" .... nc.
| |
| .y SR 3.8.4.6 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the I Inset 1
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| * design duty cycle requirements as specified in Reference 10.
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| The* Frequency of 21i months is acceptabl,*. gi-'en un' conditions;;,,, required to, pe,,,form,. the, test.and,, the, other requiroments o..t.. toesr adeuat batteryk*÷ pe...formance. durin those, 21*,. month,,* intervals. In' This SR is modified by three Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test once per 60 months. This substitution is acceptable because a modified performance discharge test represents a more severe test of battery capacity than SR 3.8.4.6. The reason for Note 2 is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems.
| |
| Due to the shared configuration of certain systems (required to mitigate DBAs and transients) between BNP Units 1 and 2, both Unit 1 and Unit 2 DC electrical power subsystems are required to supply power to these systems when either one or both units are in MODE 1, 2, or 3. In order to reduce the potential consequences associated with removing a required DC electrical power subsystem from service during the performance of (continued)
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| Brunswick Unit 1 B 3.8.4-7 Brunwic Uni No. 31 I 1 3.84-7Revision
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| | |
| DC Sources--Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.7 (continued)
| |
| REQUIREMENTS The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery performance discharge test for the duration of time equal to that of the performance discharge test.
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| 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 rate of the duty cycle). This will often confirm the battery's ability to meet the critical period 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 performance discharge test. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.7; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.7 while satisfying the requirements of SR 3.8.4.6 at the same time.
| |
| The acceptance criteria for this Surveillance is consistent with IEEE-450 (Ref. 8) and IEEE-485 (Ref. 11). These references recommend 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.
| |
| I Insert 1 -Thc Fr..qu..... for.this-tt is. norm3lly 60mnthc.,**.. If the battery shows degradation, or ifthe battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, 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. 8), when the battery capacity drops by more than 10%
| |
| relative to its capacity on the previous performance test or when it is 10%
| |
| below the manufacturer's rating. Thc 60 mont,,h Fr.....o... is co...stont
| |
| ,with*tho......... ,ion*,,, in, IEEE 150 (R,,. 8) The 12 month and 24 month Frequencies are derived from the recommendations in IEEE-450 (Ref. 8).
| |
| (continued)
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| Brunswick Unit 1 ..-
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| B 3.8.4-9 eiinN.3 No. 31 I Revision
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| | |
| Battery Cell Parameters B 3.8.6 BASES ACTIONS A.1, A.2, and A.3 (continued) parameters are restored to Category A and B limits. hee4ei Continued operation prior to declaring the affected batteries inoperable is permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable for operation prior to declaring the DC batteries inoperable.
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| B.1 When any battery parameter is outside the Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not ensured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as any Required Action of Condition A and associated Completion Time not met or average electrolyte temperature of representative cells < 60°F, also are cause for immediately declaring the associated DC electrical power subsystem inoperable.
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| SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 4), which recommends regular battery inspections (at Ic,.st onc pcr month) including voltage, specific gravity, and electrolyte temperature of pilot cells.
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| SR 3.8.6.2 '* j~Insr The etei inspection of specific gravity and voltage is consistent with IEEE-450 (Ref.4) 4). -,jlnset 1 ](continued)
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| Brunswick Unit 1 ..-
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| B 3.8.6-3 eiinN.3 No. 31 I Revision
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| | |
| Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.3 REQUIREMENTS (continued) This Surveillance verification that the average temperature of representative cells is within limits is consistent with a recommendation of IEEE-450 (Ref. 4) that states that the temperature of electrolytes in representative cells should be determined on. .... ,uatc..... basj....
| |
| .ic. **Isr 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's recommendations and the battery sizing calculations.
| |
| Table 3.8.6-1 This Table delineates the limits on electrolyte level, float voltage, and specific gravity for three different categories. The meaning of each category is discussed below.
| |
| Category A defines the normal parameter limit for each designed pilot cell in each battery. The cells selected as pilot cells are those whose temperature, voltage, and electrolyte specific gravity approximate the state of charge of the entire battery.
| |
| The Category A limits specified for electrolyte level are based on manufacturer's recommendations and are consistent with the guidance in IEEE-450 (Ref. 4), with the extra 1/4 inch allowance above the high water level indication for operating margin to account for temperature and charge effects. In addition to this allowance, Footnote (a) to Table 3.8.6-1 permits the electrolyte level to be temporarily above the specified maximum level during and following equalizing charge (i.e., for up to 3 days following the completion of an equalize charge), provided it is not overflowing. These limits ensure that the plates suffer no physical damage, and that adequate electron transfer capability is maintained in the event of transient conditions. IEEE-450 (Ref. 4) recommends that electrolyte level readings should be made only after the battery has been at float charge for at least 72 hours.
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| (continued'l Brunswick Unit 1 B 3.8.6-4 Brunwic Uni No. 31 I 1 3.86-4Revision
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| | |
| Distribution Systems--Operating B 3.8.7 BASES (continued)
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| SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. This includes verifying that distribution bus tie breakers are open and control power transfer switches associated with the 4.16 kV and 480 V emergency buses and transfer switches associated with the ESS and DG panels are aligned to their normal DC sources. The correct breaker alignment ensures the appropriate separation and independence of the electrical buses are maintained, and power is available to each required bus. The verification of energization of 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. This may be performed by verification of absence of low voltage alarms or by verifying I Inset 1
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| * a load powered from the bus is operating. The*.... 7 day,, Frequency take.....
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| into- account*.r* the redundant4*r capabi'J*lifty of* tho, AC an DC,- el'ct'**ria,-.-l per,-,*
| |
| SR 3.8.7.2 This Surveillance verifies that no combination of more than two power conversion modules (consisting of either two lighting inverters or one lighting inverter and one plant uninterruptible power supply unit) are aligned to Division II (bus B). Two power conversion modules aligned to Division II (bus B) was an initial assumption in the DC battery load study.
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| Limiting two power conversion modules to be aligned to Division II ensures the associated batteries will supply DC power to safety related equipment during a design basis event. The,*, 7*day Frequency...... , t.a, into*,,,*,,*.
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| [ Insert 1 *,* account, t*he* rounan, capab -.. **Iity of tho I'C(elect*rical per*,.. dictribution*;.
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| cubsyt... and..indicatio',+,-ns, available in the control,-, room to alen, the-REFERENCES 1. UFSAR, Chapter 6.
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| : 2. UFSAR, Chapter 15.
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| : 3. 10 CFR 50.36(c)(2)(ii).
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| Brunswick Unit 1 B3871 B3.8.7-14 eiinN.3 No. 31 I Revision
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| | |
| Distribution Systems--Shutdown B 3.8.8 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, A.2.4, and A.2.5 (continued)
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| 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 AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.
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| Notwithstanding performance of the above conservative Required Actions, a required residual heat removal-shutdown cooling (RHR-SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS would not be entered. Therefore, Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable and not in operation, which results in taking the appropriate RHR-SDC ACTIONS.
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| The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.
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| SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the AC and DC electrical power distribution subsystems are functioning properly, with the correct breaker alignment.
| |
| The correct breaker alignment ensures power is available to each required bus. The verification of energization of 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. This may be performed by verification of the absence of low voltage alarms or by
| |
| [ Inert1 l-,. verifying a load powered from the bus is operating. Th.... ,
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| 7 da Fr.qu.nc
| |
| :* ,akoc into accou..nt the reduJndanqt capability of thoe lectrical powe-r
| |
| *d;-rk ,,r-;,-ion -*cub -+,-mc, a-,s elI a"s other*, indication;,-s,- ava*'*IabeI,' in tho cotrlromtaalrthoprtrt bytomafnin.
| |
| (continued)
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| Brunswick Unit 1 B3.8.8-4 Brunwic Uni No. 31 I I 3.88-4Revision
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| | |
| Refueling Equipment Interlocks B 3.9.1 BASES (continued)
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| SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire
| |
| [Lnsrt1H* channel is tested.
| |
| Tho,- 7,,da- F......... is baed,, on engineering ju"dgmcnt and, is.
| |
| considcrcd adcquatc in '-ie'w of other indications of rcf'-cling intcr~occks and. thcir as..ociated' input statuso that÷ are avaiable to unit operations REFERENCES 1. UFSAR, Section 3.1.2.3.7.
| |
| : 2. UFSAR, Section 7.6.1.2.
| |
| : 3. UFSAR, Section 15.4.5.1.
| |
| : 4. UFSAR, Section 15.4.5.2.
| |
| : 5. 10 CFR 50.36(c)(2)(ii)
| |
| Brunswick Unit 1 B 3.9.1-4 Brunwic Uni No. 31 I 1 3.91-4Revision
| |
| | |
| Refuel Position One-Rod-Out Interlock B 3.9.2 BASES ACTIONS A.1 and A.2 (continued) all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.
| |
| SURVEILLANCE SR 3.9.2.1 REQUI REMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required.
| |
| By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in
| |
| [ Iset 1-.*. place to preclude operator errors from resulting in unanalyzed operation.
| |
| control.s<. util-..cd duri,.ng rcfu,,cing opcrations-to ..n.ure. saf operaton SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, Lnsrt1 -*,> overlapping, or total channel steps so that the entire channel is tested.
| |
| Th 7da Frqunc isc.dee adequate becaus o demo*,nstrated,,,-
| |
| circuit* reliability, procedural controls on.. +""ro rod with*drals and.I%""
| |
| visual an,.d aud,,ibled indilcatiofne avai,',lable in' the co-ntronl room' to' alertM the,-
| |
| op..rator to con.trol-, rodsI*' not-fllyF ,h in,,,ert-e,'. To perform the required testing, the applicable condition must be entered (i.e., a control rod (continued)
| |
| Brunswick Unit 1I ..-
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| B 3.9.2-3 eiinN.3 No. 31 I Revision
| |
| | |
| Control Rod Position B 3.9.3 BASES SURVEILLANCE SR 3.9.3.1 (continued)
| |
| REQUIREMENTS The, 12 hour Fro.....nc. takc,., into- co,-ns.dor.at,-,n thc procedural ,c,-ntr-,ok
| |
| _insert 1I.., ion
| |
| .... of tho, rcfu,,cli,, ;ntcrc,,.L.
| |
| REFERENCES 1. UFSAR, Section 3.1.2.3.7.
| |
| : 2. UFSAR, Section 15.4.5.1.
| |
| : 3. UFSAR, Section 15.4.5.2.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1B3.33ReionN.1 B 3.9.3-3 Revision No. 31
| |
| | |
| Control Rod OPERABILITY--Refueling B 3.9.5 BASES SURVEILLANCE SR 3.9.5.1 and SIR 3.9.5.2 (continued)
| |
| REQUIREMENTS and the associated CR0 scram accumulator pressure is ->940 psig.
| |
| SR 3.9.5.2 may be performed by verification of absence of the common scram accumulator low pressure alarm.
| |
| Thc, 7 day Frcguency takes, into con..ideration c....me. re...{Jablityhl,,
| |
| SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3.
| |
| REFERENCES 1. UFSAR, Section 3.1.2.3.7.
| |
| : 2. UFSAR, Section 15.4.5.1.
| |
| : 3. UFSAR, Section 15.4.5.2.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B 3.9.5-3 BrunwicUni 1 3.95-3Revision No. 31 ]
| |
| | |
| RPV Water Levei B 3.9.6 BASES (continued)
| |
| LCO A minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits.
| |
| APPLICABILITY LCO 3.9.6 is applicable when moving fuel assemblies or handling control rods (i.e., movement with other than the normal control rod drive) within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the RPV, there can be no significant radioactivity release as a result of a postulated fuel handling accident.
| |
| Requirements to preclude fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.7, "Spent Fuel Storage Pool Water Level ."
| |
| ACTIONS A.1 If the water level is < 23 ft above the top of irradiated fuel assemblies seated within the RPV, all operations involving movement of fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and control rod handling shall not preclude completion of movement of a component to a safe position.
| |
| SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of irradiated fuel assemblies seated within the RPV ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in
| |
| [ Inset 1
| |
| * containment (Ref. 2).
| |
| Thc Freque*,.ncy, of 21 hour,-e is- bacdo cgcrngjug,,4ndi con.drc..... ci vewo thc"arg volm.c;,', of wtc.. nd th normdl proccdural control-"s on valve- position;,"* which makc significant* unplanned (continued)
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| Brunswick Unit 1 ..-
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| B3.g.6-2 eiinN.3 No. 31 I Revision
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| | |
| RHR--High Water Level B 3.9.7 BASES ACTIONS B.1, B.2, B.3, and B.4 (continued) examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, a Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
| |
| C.1 and 0.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour.
| |
| The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
| |
| Furthermore, verification of reactor coolant circulation must be reconfirmed every 12 hours thereafter. This will ensure reactor coolant circulation is maintained.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| SURVEILLANCE SR 3.9.7.1 REQUIREMENTS This Surveillance demonstrates that the required RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient I Iser-I decay heat removal capability.
| |
| -Tho Frcqucnc'; of 12, hor i...cuf.... n in.,,* vio, of;o...t
| |
| ,hr... ....u. n'd audiblo.
| |
| ;indication; a.. il.bl to*k, t ho opcr... for*"monitoring tho" RHR ..hutdo.n REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1 B3.9.7-4 Brunwic Uni I No. 31 I 3.97-4Revision
| |
| | |
| RHR--Low Water Level B 3.9.8 BASES ACTIONS B.1, B.2. and B.3 (continued)
| |
| If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
| |
| 0.1 and 0.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour.
| |
| The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation.
| |
| Furthermore, verification of reactor coolant circulation must be reconfirmed every 12 hours thereafter. This will ensure reactor coolant circulation is maintained.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| SURVEILLANCE SR 3.9.8.1 REQUI REMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient I Inert*...,,decay heat removal capability.
| |
| Thc-Fr......... of 12 hou.... su...ficiont, in. VioW of* othc.i.u .. l nd au-d,
| |
| ,,;ib-indications available to thc opcrator for menitorin~g the RHR. shutdown REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit 1B398-ReionN.1 B 3.9.8-4 Revision No. 31 I
| |
| | |
| Reactor Mode Switch Interlock Testing B 3.10.2 BASES SURVEILLANCE SR 3.10.2.1 and SR 3.10.2.2 (continued)
| |
| REQUI REMENTS The administrative controls are to be periodically verified to ensure that the operational requirements continue to be met. In addition, the all rods fully inserted Surveillance (SR 3.10.2.1) must be verified by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other technically qualified member of the unit technical staff. A member of the technical staff is considered to be qualified if he has completed applicable qualification requirements in accordance with required plant training and qualification procedures. Thc Sur-cillances porformed at tho 12 hour and 21 hour Frequencie. .. re intended,,, topoieaprpit,,srneta each.oprtn.. thosc hi.. is.... are..of.. nd....rifie. compliance wit*h...
| |
| REFERENCES 1. UFSAR, Section 7.2.1.1.2.13.
| |
| : 2. UFSAR, Section 15.4.5.1.
| |
| : 3. UFSAR, Section 15.4.5.2.
| |
| : 4. 10 CFR 50.36(c)(2)(ii).
| |
| Brunswick Unit I B 3.10.2-5 BrunwickUni I B3.102-5Revision No. 31
| |
| | |
| Single Control Rod Withdrawal--Hot Shutdown B 3.10.3 BASES ACTIONS A..__ (continued) rods. This Required Action includes exiting this Special Operations Applicability by returning the reactor mode switch to the shutdown position. A second Note has been added, which clarifies that this Required Action is only applicable if the requirements not met are for an affected LCO.
| |
| A.2.1 and A.2.2 Required Actions A.2.1 and A.2.2 are alternate Required Actions that can be taken instead of Required Action A.1 to restore compliance with the normal MODE 3 requirements, thereby exiting this Special Operations LCO's Applicability. Actions must be initiated immediately to insert all insertable control rods. Actions must continue until all such control rods are fully inserted. Placing the reactor mode switch in the shutdown position will ensure all inserted rods remain inserted and restore operation in accordance with Table 1.1-1. The allowed Completion Time of 1 hour to place the reactor mode switch in the shutdown position provides sufficient time to normally insert the control rods.
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| SURVEILLANCE SR 3.10.3.1. SR 3.10.3.2. and SR 3.10.3.3 REQUIREMENTS The other LCOs made applicable in this Special Operations LCO are required to have their Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed while the scram function for the withdrawn rod is not available, periodic verification in accordance with SR 3.10.3.2 is required to preclude the possibility of criticality. SR 3.10.3.2 has been modified by a Note, which clarifies that this SR is not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements, since SR 3.10.3.2 demonstrates that the alternative LCO 3.10.3.d.2 requirements are satisfied. Also, SR 3.10.3.3 verifies that all control rods other than the control rod being withdrawn are fully inserted. Tho, 21A hou Froquoncy,,i.. .,
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| acopalobcacco thmc adlminmtatvct controvlls on* controml rod (continued)
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| Brunswick Unit 1 B31.-
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| B 3.10.3-4 eiinN.3 No. 31 I Revision
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| Single Control Rod Withdrawal--Hot Shutdown B 3.10.3 BASES SURVEILLANCE SR 3.10.3.1, SR 3.10.3.2, and SR 3.10.3.3 (continued)
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| REQUIREMENTS REFERENCES 1. UFSAR, Section 15.4.5.1.
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| : 2. 10 CFR 50.36(c)(2)(ii).
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| Brunswick Unit 1 B 3.10.3-5 No. 31 I BrunwickUni 1 B3.103-5Revision
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| Single Control Rod Withdrawal--Cold Shutdown B 3.10.4 BASES (continued)
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| SURVEILLANCE SR 3.10.4.1. SR 3.10.4.2. SR 3.10.4.3. and SR 3.10.4.4 REQUIREMENTS The other LCOs made applicable by this Special Operations LCO are required to have their associated surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and electrically disarmed while the scram function for the withdrawn rod is not available, periodic verification is required to ensure that the possibility of criticality remains precluded. Verification that all the other control rods are fully inserted is required to meet the 3DM requirements. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the affected control rod. The 2hour Frequenc....... is wvithdr..al.. the pro..f.-,.~tecio aFffordod* by,thc' IC~ in
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| *VOlVed, and herlardwir SR 3.10.4.2 and SR 3.10.4.4 have been modified by Notes, which clarify that these SRs are not required to be met ifthe alternative requirements demonstrated by SR 3.10.4.1 are satisfied.
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| REFERENCES 1. UFSAR, Section 15.4.5.
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| : 2. 10 CFR 50.36(c)(2)(ii).
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| Brunswick Unit 1 B31.-
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| B 3.10.4-5 eiinN.3 No. 31 I Revision
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| Single CR0 Removal--Refueling B 3.10.5 BASES (continued)
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| ACTIONS A.1, A.2.1. and A.2.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for failure to meet LCO 3.3.1.1, LCO 3.9.1, LCO 3.9.2, LCO 3.9.4, and LCO 3.9.5 (i.e., all control rods inserted) or with the allowances of this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2.1, and Required Action A.2.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the CR0 and insert its control rod, or initiate action to restore compliance with this Special Operations LCO. Actions must continue until either Required Action A.2.1 or Required Action A.2.2 is satisfied.
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| SURVEILLANCE SR 3.10.5.1. SR 3.10.5.2. SR 3.10.5.3, SR 3.10.5.4, and SR 3.10.5.5 REQUIREMENTS Verification that all the control rods, other than the control rod withdrawn for the removal of the associated CR0, are fully inserted is required to ensure the SDM is within limits. Verification that the local five by five array of control rods, other than the control rod withdrawn for removal of the associated CR0, is inserted and electrically disarmed, while the scram function for the withdrawn rod is not available, is required to ensure that the possibility of criticality remains precluded. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the withdrawn control rod. The Surveillance for LCO 3.1.1, which is made applicable by this Special Operations LCO, is required in order to establish that this Special Operations LCO is being met. Verification that no other CORE ALTERATIONS are being made is required to ensure the assumptions of the safety analysis are satisfied.
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| Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an Iner 2inadvertent criticality. Thc 21, hou-r Frcqucncy i...3ccc...... given tho
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| * dministrati-e controls on, control rod remo'-al and hardw-rc intorlock to (continued)
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| Brunswick Unit 1 BB31.-
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| 3.10.5-4 eiinN.3 No. 31 [
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| Revision
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| Multiple Control Rod Withdrawal--Refueling B 3.10.6 BASES APPLICABILITY appropriately controlled by requiring all fuel to be removed from cells (continued) whose "full-in" indicators are allowed to be bypassed.
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| ACTIONS A.1, A.2. A.3.1. and A.3.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required ActionS restores operation consistent with the normal requirements for refueling (i.e., all control rods inserted in core cells containing one or more fuel assemblies) or with the exceptions granted by this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2, Required Action A.3.1, and Required Action A.3.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the affected CRDs and insert their control rods, or initiate action to restore compliance with this Special Operations LCO.
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| SURVEILLANCE SR 3.10.6.1. SR 3.10.6.2, and SR 3.10.6.3 REQUIREMENTS Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an
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| [ Insert 2
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| * inadvertent criticality. Th,, 2, hou
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| .... cq....... i"..... ptbl,,, gie
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| ... th,,
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| REFERENCES 1. UFSAR, Section 15.4.5.
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| : 2. 10 CFR 50.36(c)(2)(ii).
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| Brunswick Unit 1 BB31.-
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| 3.10.6-3 eiinN.3 No. 31 I Revision
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| SDM Test-Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.1. SR 3.10.8.2. and SR 3.10.8.3 (continued)
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| REQUIREMENTS However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff. As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These surveillances provide adequate assurance that the specified test sequence is being followed.
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| SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full-out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
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| SR 3.10.8.6 CR0 charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is (continued)
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| Brunswick Unit 1 B 3.10.8-5 BrunwickUniNo. 31 I I B3.108-5Revision
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| 3DM Test--Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.6 (continued)
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| REQUIREMENTS insufficient reactor pressure to scram the control rods. Verification of charging water header pressure ensures that if a scram were required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig is well below the expected pressure of approximately 1400 psig while still ensuring sufficient pressure for rapid i Insert 1
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| * control rod insertion. Th,, 7 day Fro u....... h,, bc,
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| .... n to..b,,,
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| REFERENCES 1. UFSAR, Section 15.4.6.
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| : 2. 10 CFR 50.36(c)(2)(ii).
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| Brunswick Unit 1 B 3.10.8-6 BrunwickUni I B3.108-6Revision No. 31
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| BSEP 15-0101 Enclosure 6 Page 1 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description 1_]TSTF-425_ Brunswick Reactivity Anomalies N/A 3.1.2 Verify core reactivity N/A 3.1.2.1 Control Rod OPERABILITY 3.1.3 3.1.3 Determine control rod position 3.1.3.1 3.1.3.1 Perform notch test - fully withdrawn control rods 3.1.3.2 N/A Perform notch test - withdrawn control rods 3.1.3.3 3.1.3.2 Control Rod Scram Times 3.1.4 3.1.4 Perform scram time testing 3.1.4.2 3.1.4.2 Control Rod Scram Accumulators 3.1.5 3.1.5 Verify control rod scram accumulator pressure 3.1.5.1 3.1.5.1 Rod Pattern Control 3.1.6 3.1.6 Verify control rods comply with withdrawal sequence 3.1.6.1 3.1.6.1 Standby Liquid Control System 3.1.7 3.1.7 Verify volume of sodium pentaborate solution 3.1.7.1 3.1 .7.1 Verify temperature of sodium pentaborate solution 3.1.7.2 3.1.7.2 Verify temperature of SLC piping 3.1.7.3 3.1.7.3 Verify continuity of explosive charge 3.1.7.4 3.1.7.4 Verify concentration of boron solution 3.1.7.5 3.1.7.5 Verify manual/power operated valve positon 3.1.7.6 N/A Verify pump flow rate 3.1.7.7 3.1.7.6 2 Verify flow through SLC subsystem 3.1.7.8 3.1.7.7 Verify heat traced piping is unblocked 3.1.7.9 N/A Scram Discharge Volume (SDV) Vent and Drain Valves 3.1.8 3.1.8 Verify each SDV vent and drain valve is open 3.1.8.1 3.1.8.1 Cycle each SDV vent and drain valve 3.1.8.2 3.1.8.2 Verify automatic operation of each SDV vent and drain valve 3.1.8.3 3.1.8.3 Average Planar Linear Heat Generation Rate (APLHGR) 3.2.1 ' 3.2.1 Verify all APLHGRs are less than or equal to limits 3.2.1.1 3.2.1.1 Minimum Critical Power Ratio (MCPR) 3.2.2 3.2.2 Verify all MCPRs are greater than or equal to limits 3.2.2.1 3.2.2.1 Linear Heat Generation Rate (LHGR) 3.2.3 3.2.3 Verify all LHGRs are less than or equal to limits 3.2.3.1 3.2.3.1 Average Power Range Monitor Gain and Setpoints 3.2.4 N/A
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| 'Verify MFLPD is within limits 3.2.4.1 N/A
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| BSEP 15-0101 Enclosure 6 Page 2 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description 1 TSTF-425 Brunswick Verify APRM setpoints or gains are adjusted for the calculated 3.2.4.2 N/A MFLPD Reactor Protection System (RPS) Instrumentation 3.3.1.1 3.3.1.1 Perform CHANNEL CHECK 3.3.1.1.1 3.3.1.1.2 Verify absolute difference between APRM channels and 3.3.1.1.2 N/A calculated power Adjust the channel to conform to a calibrated flow signal 3.3.1.1.3 N/A Adjust the APRM channels to conform to the calculated power N/A 3.3.1.1.3 Perform CHANNEL FUNCTIONAL TEST 3.3.1.1.4 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST 3.3.1.1.5 N/A Perform a functional test of each automatic scram contactor N/A 3.3.1.1.5 Verify the IRM and APRM channels overlap N/A 3.3.1.1.7 Calibrate the local power range monitors 3.3.1.1.6 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST 3.3.1.1.7 3.3.1.1.9 Calibrate the trip units 3.3.1.1.8 3.3.1.1.10 Perform CHANNEL CALIBRATION 3.3.1.1.9 N/A Perform CHANNEL FUNCTIONAL TEST N/A 3.3.1.1.11 Perform CHANNEL FUNCTIONAL TEST 3.3.1.1.10 3.3.1.1.12 Perform CHANNEL CALIBRATION 3.3.1.1.11 3.3.1.1.13 Verify the APRM Flow Biased Simulated Thermal Power 3.3.1.1.12 N/A Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.1.1.13 3.3.1.1.15 Verify Turbine Stop Valve - Closure and Turbine Control Valve 3.3.1.1.14 3.3.1.1.16 Fast Closure, Trip Oil Pressure - Low Functions are not bypassed Verify the RPS RESPONSE TIME is within limits 3.3.1.1.15 3.3.1.1.17 Verify OPRM is not bypassed when APRM Simulated Thermal N/A 3.3.1 .1 .19 Power is > 25% and recirculation drive flow is < 60%
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| Source Range Monitor (SRM) Instrumentation 3.3.1.2 3.3.1.2 Perform CHANNEL CHECK 3.3.1.2.1 3.3.1.2.1 Verify an OPERABLE SRM detector 3.3.1.2.2 3.3.1.2.2 Perform CHANNEL CHECK 3.3.1.2.3 3.3.1.2.3 Verify count rate 3.3.1.2.4 3.3.1.2.4 Perform CHANNEL FUNCTIONAL TEST 3.3.1.2.5 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST 3.3.1.2.6 3.3.1.2.6 Perform CHANNEL CALIBRATION 3.3.1.2.7 3.3.1.2.7
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| BSEP 15-01 01 Enclosure 6 Page 3 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description 1 [TSTF-425_]Brunswick Control Rod Block Instrumentation 3.3.2.1 3.3.2.1 Perform CHANNEL FUNCTIONAL TEST 3.3.2.1.1 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST 3.3.2.1.2 3.3.2.1.2 Perform CHANNEL FUNCTIONAL TEST 3.3.2.1.3 3.3.2.1.3 Verify the RBM 3.3.2.1.4 3.3.2.1.4 Verify the RWM is not bypassed 3.3.2.1 .5 3.3.2.1 .5 Perform CHANNEL FUNCTIONAL TEST 3.3.2.1.6 3.3.2.1.6 Perform CHANNEL CALIBRATION 3.3.2.1.7 3.3.2.1.7 Feedwater and Main Turbine High Water Level Trip 3.3.2.2 3.3.2.2 Instrumentation Perform CHANNEL CHECK 3.3.2.2.1 3.3.2.2.1 Perform CHANNEL FUNCTIONAL TEST 3.3.2.2.2 N/A Perform CHANNEL CALIBRATION 3.3.2.2.3 3.3.2.2.2 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.2.2.4 3.3.2.2.3 Post Accident Monitoring (PAM) Instrumentation 3.3.3.1 3.3.3.1 Perform CHANNEL CHECK 3.3.3.1.1 3.3.3.1.1 Perform CHANNEL CALIBRATION 3.3.3.1.2 3.3.3.1.3 Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3.3.2 Perform CHANNEL CHECK 3.3.3.2.1 3.3.3.2.1 Verify each required control circuit and transfer switch 3.3.3.2.2 N/A Perform CHANNEL CALIBRATION 3.3.3.2.3 3.3.3.2.2 End of Cycle Recirculation Pump Trip (EOC-RPT) 3.3.4.1 N/A Instrumentation Perform CHANNEL FUNCTIONAL TEST 3.3.4.1.1 N/A Calibrate trip units 3.3.4.1.2 N/A Perform CHANNEL CALIBRATION 3.3.4.1.3 N/A Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.4.1.4 N/A Ver'ify TSV - Closure and TCV Fast Closure, Trip Oil Pressure - 3.3.4.1 .5 N/A Low Functions are not bypassed Verify the EOC-RPT SYSTEM RESPONSE TIME 3.3.4.1.6 N/A Determine RPT breaker interruption time 3.3.4.1.7 N/A Anticipated Transient Without Scram Recirculation Pump Trip 3.3.4.2 3.3.4.1 (ATWS-RPT) Instrumentation Perform CHANNEL CHECK 3.3.4.2.1 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST 3.3.4.2.2 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST 3.3.4.2.2 3.3.4.1.2
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| BSEP 15-0101 Enclosure 6 Page 4 of 16 Technical Specification Cross Reference for TSTF-425 and BrunswickMark-ups Technical Specification Section Title/Surveillance Description 1 _TSTF-425_jBrunswick Calibrate the trip unit 3.3.4.2.3 3.3.4.1 .3 Perform CHANNEL CALIBRATION 3.3.4.2.4 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.4.2.5 3.3.4.1.5 Emergency Core Cooling System (ECCS) Instrumentation 3.3.5.1 3.3.5.1 Perform CHANNEL CHECK 3.3.5.1.1 3.3.5.1.1 Perform CHANNEL FUNCTIONAL TEST 3.3.5.1.2 3.3.5.1.2 Calibrate the trip unit 3.3.5.1.3 3.3.5.1.3 Perform CHANNEL CALIBRATION 3.3.5.1.4 N/A Perform CHANNEL CALIBRATION 3.3.5.1.5 3.3.5.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.5.1.6 3.3.5.1.5 Perform CHANNEL FUNCTIONAL TEST N/A 3.3.5.1.6 Verify the ECCS RESPONSE TIME is within limits 3.3.5.1.7 N/A Reactor Core Isolation Cooling (RClC) System Instrumentation 3.3.5.2 3.3.5.2 Perform CHANNEL CHECK 3.3.5.2.1 3.3.5.2.1 Perform CHANNEL FUNCTIONAL TEST 3.3.5.2.2 3.3.5.2.2 Calibrate the trip units 3.3.5.2.3 3.3.5.2.3 Perform CHANNEL CALIBRATION 3.3.5.2.4 N/A Perform CHANNEL CALIBRATION 3.3.5.2.5 3.3.5.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.5.2.6 3.3.5.2.5 Primary Containment Isolation Instrumentation 3.3.6.1 3.3.6.1 Perform CHANNEL CHECK 3.3.6.1.1 3.3.6.1.1 Perform CHANNEL FUNCTIONAL TEST 3.3.6.1.2 3.3.6.1.2 Calibrate the trip unit 3.3.6.1.3 3.3.6.1.3 Perform CHANNEL CALIBRATION 3.3.6.1.4 3.3.6.1.4 Perform CHANNEL FUNCTIONAL TEST 3.3.6.1.5 3.3.6.1.5 Perform CHANNEL CALIBRATION 3.3.6.1.6 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.6.1.7 3.3.6.1.7 Verify the ISOLAITON SYSTEM RESPONSE TIME is within 3.3.6.1.8 3.3.6.1.8 limits Perform CHANNEL FUNCTIONAL TEST N/A 3.3.6.1.9 Secondary Containment Isolation Instrumentation 3.3.6.2 3.3.6.2 Perform CHANNEL CHECK 3.3.6.2.1 3.3.6.2.1 Perform CHANNEL FUNCTIONAL TEST 3.3.6.2.2 3.3.6.2.2 Calibrate the trip unit 3.3.6.2.3 3.3.6.2.3 Perform CHANNEL CALIBRATION 3.3.6.2.4 N/A
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| BSEP 15-01 01 Enclosure 6 Page 5 of 16 Technical Specification Cross Reference for TSTF-425 and BrunswickMark-ups Technical Specification Section Title/Surveillance Description' TSTF-425 Brunswick Perform CHANNEL CALIBRATION 3.3.6.2.5 3.3.6.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.6.2.6 3.3.6.2.5 Verify the ISOLAITON SYSTEM RESPONSE TIME is within 3.3.6.2.7 N/A limits Low-Low Set (LLS) Instrumentation , ,3.3.6.3 NIA Perform CHANNEL CHECK 3.3.6.3.1 N/A Perform CHANNEL FUNCTIONAL TEST 3.3.6.3.2 N/A Perform CHANNEL FUNCTIONAL TEST 3.3.6.3.3 N/A Perform CHANNEL FUNCTIONAL TEST 3.3.6.3.4 N/A Calibrate the trip unit 3.3.6.3.5 N/A Perform CHANNEL CALIBRATION 3.3.6.3.6 N/A Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.6.3.7 N/A Main Control Room Environmental Control (MCREC) System 3..3.7.1 3.3.7.1 instrumentation / Control Room Emergency Ventilation (CREV) System Instrumentation ..
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| Perform CHANNEL CHECK 3.3.7.1.1 3.3.7.1.1 Perform CHANNEL FUNCTIONAL TEST 3.3.7.1.2 3.3.7.1.2 Calibrate the trip unit 3.3.7.1.3 N/A Perform CHANNEL CALIBRATION 3.3.7.1.4 3.3.7.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.7.1.5 3.3.7.1.4
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| =Condenser Vacuum Pump isolation Instrumentation i iN/A . 3..
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| r3.7.2 ..
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| Perform CHANNEL CHECK N/A 3.3.7.2.1 Perform CHANNEL FUNCTIONAL TEST N/A 3.3.7.2.2 Perform CHANNEL CALIBRATION N/A 3.3.7.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST N/A 3.3.7.2.4 LOss of Power (LoP) !nstrumentation 3.3.8.1 3.3.8.1 Perform CHANNEL CHECK 3.3.8.1.1 N/A Perform CHANNEL FUNCTIONAL TEST 3.3.8.1.2 3.3.8.1.1 Perform CHANNEL CALIBRATION 3.3.8.1 .3 3.3.8.1 .2 Perform CHANNEL CALIBRATION N/A 3.3.8.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST 3.3.8.1.4 3.3.8.1.4 Reactor Protection System (RPS) Electric Power Monitoring 3.3.8.2 3.3.8.2 Perform CHANNEL FUNCTIONAL TEST 3.3.8.2.1 3.3.8.2.1 Perform CHANNEL CALIBRATION 3.3.8.2.2 3.3.8.2.2 Perform CHANNEL CALIBRATION N/A 3.3.8.2.3
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| BSEP 15-01 01 Enclosure 6 Page 6 of 16 Technical Specification Cross Reference for TSTF-425 and BrunswickMark-ups Technical Specification Section Title/Surveillance Description' _TSTF-425 Brunswick Perform a system functional test 3.3.8.2.3 3.3.8.2.4 Recirculation Loops Operating ' 3.4.1 3.4.1 Verify recirculation loop jet pump flow mismatch 3.4.1.1 3.4.1.1
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| -Jet Pumps" , 3.4.2 3.4.2 Verify at least one criteria is satisfied for each operating 3.4.2.1 3.4.2.1 recirculation loop Safety/Relief Valves (SRvs) 3.4.3 3.4.3 Verify the safety function lift setpoints of the SRVs 3.4.3.1 3.4.3.1 2 Verify each required SRV opens when manually actuated 3.4.3.2 3.4.3.2 RCS0Operational LEAKAGE .... 3.4.4 3,4.4 Verify RCS unidentified and total LEAKAGE and unidentified 3.4.4.1 3.4.4.1 LEAKAGE increase are within limits iC~S Pressure Isolation valve (PIV) Leakage 34.N/
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| Verify equivalent leakage of each RCS PIV 3.4.5.1 N/A RCS Leakage Detection Instrumentation . .. 3.4.8 3.4,5*
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| Perform a CHANNEL CHECK ______________3.4.6.1 3.4.5.1 Perform CHANNEL FUNCTIONAL TEST 3.4.6.2 3.4.5.2 Perform CHANNEL CALIBRATION 3.4.6.3 3.4.5.3 RCS Specific Activity . .. 347* 3.4.6 Verify reactor coolant DOSE EQUIVALENT I-131 3.4.7.1 3.4.6.1 Residual Heat Removal (RHR) Shutdown Cooling System -Hot 3.4.8 3.4.7 Shutdown . ..... .. . .
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| Verify one required RHR shutdown cooling subsystem or 3.4.8.1 3.4.7.1 recirculation pump is operating Residual Heat Removal (RHR) Shutdown Cooling System - ,3.4.9 3.4.8 Cold Shutdown Verify one required RHR shutdown cooling subsystem or 3.4.9.1 3.4.8.1 recirculation pump is operating______
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| RCS Pressure and Temperature (P/T) Limits 3.4.10 3.4.9 "
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| Verify RCS pressure, temperature, heatup/cooldown N/A 3.4.9.1 (heatup/cooldown operations)
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| Verify RCS pressure, temperature, heatup/cooldown (inservice 3.4.10.1 3.4.9.2 leak and hydrostatic testing)
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| Verify reactor vessel flange and head flange temperatures 3.4.10.7 3.4.9.6 (tensioning)
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| Verify reactor vessel flange and head flange temperatures 3.4.10.8 3.4.9.7 (Mode 4 80 degrees F) ___________
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| BSEP 15-0101 Enclosure 6 Page 7 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description' _TSTF-425_JBrunswick Verify reactor vessel flange and head flange temperatures 3.4.10.9 3.4.9.8 (Mode 4 100 degrees F)
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| Reactor Steam Dome Pressure 3.4.11 3.4.10 Verify reactor steam dome pressure 3.4.11.1 3.4.10.1 ECCS - Operating 3.5.1 3.5.1 Verify each EGOS injection/spray subsystem piping is filled with 3.5.1.1 3.5.1.1 water Verify each EGOS injection/spray subsystem manual, power 3.5.1.2 3.5.1.2 operated, and automatic valve is in the correct position Verify ADS air supply header pressure 3.5.1.3 3.5.1.3 Verify the RHR System cross tie valve is locked closed 3.5.1.4 3.5.1.4 Verify each LPCI inverter output voltage 3.5.1 .5 N/A Verify ECOS pumps flow rates 3.5.1.7 3.5.1.6 Verify HPCI pump flow rate 3.5.1.8 3.5.1.7 Verify HPCI pump flow rate 3.5.1.9 3.5.1.8 Verify each EGGS injection/spray subsystem automatic initiation 3.5.1.10 3.5.1.9 Verify the ADS automatic initiation 3.5.1.11 3.5.1.10 Verify each required ADS valve opens when manually actuated 3.5.1.12 3.5.1.11 Verify the EGOS RESPONSE TIME for each required EGOS N/A 3.5.1.12 injection/spray subsystem ECCS - Shutdown 3.5.2 3.5.2 Verify the suppression pool water 3.5.2.1 3.5.2.1 Verify, for each required core spray (CS) subsystem, 3.5.2.2 3.5.2.2 suppression pool water level and condensate storage tank water level/volume Verify piping is filled with water 3.5.2.3 3.5.2.3 Verify manual, power operated, and automatic valve position 3.5.2.4 3.5.2.4 Verify each required EGOS pump develops the specified flow 3.5.2.5 3.5.2.5 Verify each required EGOCS injection/spray subsystem automatic 3.5.2.6 3.5.2.6 initiation Verify the EGOS RESPONSE TIME for each required EGOS N/A 3.5.2.7 injection/spray subsystem RClC System 3.5.3 3.5.3 Verify the ROIG System piping is filled with water 3.5.3.1 3.5.3.1 Verify each ROIG System manual, power operated, and 3.5.3.2 3.5.3.2 automatic valve position ____________
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| Verify the RCIC pump flow rate 3.5.3.3 3.5.3.3
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| BSEP 15-0101 Enclosure 6 Page 8 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description' [TSTF-425_ Brunswick Verify the RCIC pump flow rate 3.5.3.4 3.5.3.4 Verify the RCIC System automatic initiation 3.5.3.5 3.5.3.5 Primary ontainment 3.6.1.1 3.6.1.1 Verify drywell to suppression chamber differential pressure drop 3.6.1 .1 .2 3.6.1.1.2 Primary Containment Air Lock '*... ... 3.6.1.2 3.6.1,2 Verify only one door in the primary containment air lock can be 3.6.1.2.2 3.6.1.2.2 opened at a time Primary Containment Isolation Valves (PClVs) 3.6,1,3 3.6.1.3 Verify each primary containment purge valve is sealed closed 3.6.1.3.1 N/A except one Verify each primary containment purge valve is closed 3.6.1.3.2 N/A Verify each primary containment isolation manual valve and 3.6.1.3.3 3.6.1.3.1 blind flange that is located outside primary containment is closed.
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| Verify continuity of the traversing incore probe (TIP) shear 3.6.1.3.5 3.6.1.3.3 isolation valve explosive charge Verify the isolation time of each power operated and each 3.6.1.3.6 3.6.1.3.4 2 automatic PCIV, except for MSIVs, is within limits Perform leakage rate testing for each primary containment 3.6.1.3.7 N/A purge valve with resilient seals Verify the isolation time of each MSIV 3.6.1.3.8 3.6.1.3.5 2 Verify each automatic PCIV actuates to the isolation position 3.6.1.3.9 3.6.1.3.6 Verify a representative sample of reactor instrumentation line 3.6.1.3.10 3.6.1.3.7 EEFCVs actuate Remove and test the explosive squib from each shear isolation 3.6.1.3.11 3.6.1.3.8 2 valve of the TIP System Verify each primary containment purge valve is blocked to 3.6.1.3.15 N/A restrict valve from opening______
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| DryWell Pressure 3.6.1.4' N/A Verify drywell pressure is within limits 3.6.1.4.1 N/A Drywell Air Temperature * . .3.6.1.5 3.6.1 .4 Verify drywell average air temperature is within limit 3.6.1.5.1 3.6.1.4.1 Low-Low Set (LLS) Valves .. . .. 3.6.1.6 N/A Verify each LLS valve opens when manually actuated 3.6.1.6.1 N/A Verify the LLS System automatic initiation 3.6.1.6.2 N/A
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| BSEP 15-0101 Enclosure 6 Page 9 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description 1 _TSTF-425_ Brunswick Reactor Building-to-Suppression Chamber Vacuum Breakers~i 3.6.1.7* 3.6.1.5...
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| Verify nitrogen bottle supply pressure of each nitrogen backup N/A 3.6.1.5.1 subsystem Verify each vacuum breaker is closed 3.6.1.7.1 3.6.1.5.2 Perform a functional test of each vacuum breaker 3.6.1.7.2 3.6.1.5.3 Verify the open setpoint of each vacuum breaker 3.6.1.7.3 3.6.1.5.4 Verify leakage rate of each nitrogen backup subsystem N/A 3.6.1.5.5 Verify the Nitrogen Backup System supplies nitrogen to the N/A 3.6.1.5.6 vacuum breakers on an actuation signal Suppression Chamber-to-Drywell Vacuum Breakers '3.6.1.8 ... 3.6.1.6 Verify each vacuum breaker is closed 3.6.1.8.1 3.6.1.6.1 Perform a functional test of each required vacuum breaker 3.6.1.8.2 3.6.1.6.2
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| *Verify the full open setpoint of each vacuum breaker 3.6.1.8.3 3.6.1.6.3 Main steam Isolation Valve (MSIV) Leakage Control System. 3.6.1.9 ... N/A Operate each MSIV LOS blower 3.6.1.9.1 N/A Verify electrical continuity of each inboard MSIV LOS subsystem 3.6.1.9.2 N/A heater element circuitry Perform a system functional test of each MSIV LOS subsystem 3.6.1.9.3 N/A Suppression Pool Average Temperature 3.6.2.1 3.6,2.1 Verify suppression pool average temperature is within limits 3.6.2.1.1 3.6.2.1.1 Suppression Pool Water Level 3.6.2.2 3.6.2.2 Verify suppression pool water level is within limits 3.6.2.2.1 3.6.2.2.1 Residual Heat Removal (RHR) Suppression Pool Cooling .... 3.6.?.3 3.6.2.3 Verify each RHR suppression pool cooling subsystem manual, 3.6.2.3.1 3.6.2.3.1 power operated, and automatic valve is in the correct position or can be aligned to the correct position Verify each RHR pump a flow rate 3.6.2.3.2 3.6.2.3.2 Residual Heat Removal (RHR) Suppression Pool Spray 3.6.2.4 N/A ..
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| Verify each RHR suppression pool cooling subsystem manual, 3.6.2.4.1 N/A power operated, and automatic valve is in the correct position or can be aligned to the correct position Verify each RHR pump a flow rate 3.6.2.4.2 N/A Dryweil-t0-Suppression Chamber Differential Pressure 3.6.2.5 N/A Verify drywell-to-suppression chamber differential pressure is 3.6.2.5.1 N/A w ithin lim its ____________________ ___________________
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| BSEP 15-0101 Enclosure 6 Page 10 of 16 Technical Specification Cross Reference for TSTF-425 and BrunswickMark-ups Technical Specification Section Title/Surveillance Description 1 TSTF-425[ Brunswick Drywell Cooling System Fans 3.6.3.1 N/A Operate each required drywell cooling system fan 3.6.3.1.1 N/A Verify each required drywell cooling system fan flow rate 3.6.3.1.2 N/A Primary Containment Oxygen Concentration 3.6.3.2 3.6.3.1 Verify primary containment oxygen concentration is within limits 3.6.3.2.1 3.6.3.1 .1 Containment Atmosphere Dilution (CAD) System 3.6.3.3 N/A Verify volume of liquid nitrogen 3.6.3.3.1 N/A Verify each CAD subsystem manual, power operated, and 3.6.3.3.2 N/A automatic valve is in the correct position or can be aligned to the correct position Secondary Containment 3.6.4.1 3.6.4.1 Verify secondary containment vacuum 3.6.4.1.1 N/A Verify all secondary containment equipment hatches are closed 3.6.4.1.2 3.6.4.1.1 and sealed Verify one secondary containment access door is closed in each 3.6.4.1.3 3.6.4.1.2 access opening Verify secondary containment can be drawn down using one 3.6.4.1 .4 N/A SGT subsystem______
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| Verify each SGT subsystem can maintain vacuum water gauge 3.6.4.1.5 3.6.4.1 .3 in the secondary containment for 1 hour Secondary Containment Isolation Valves (SCIVs) / Secondary 3.6.4.2 3.6.4.2 Containment Isolation Dampers (SCIDs)
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| Verify each secondary containment isolation manual valve and 3.6.4.2.1 N/A blind flange that is required to be closed during accident conditions is closed.
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| Verify the isolation time of each automatic SCIV is within limits 3.6.4.2.2 3.6.4.2.1 Verify each automatic SCIV actuates to the isolation position 3.6.4.2.3 3.6.4.2.2 Standby Gas Treatment (SGT) System 3.6.4.3 3.6.4.3 Operate each SGT subsystem with heaters operating 3.6.4.3.1 3.6.4.3.1 Verify each SGT subsystem actuates 3.6.4.3.3 3.6.4.3.3 Verify each SOT filter cooler bypass damper can be opened and 3.6.4.3.4 N/A the fan started Residual Heat Removal Service Water (RHRSW) System 3.7.1 3.7.1 Verify each RHRSW manual, power operated, and automatic 3.7.1.1 3.7.1.1 valve is in the correct position or can be aligned to the correct position____ ___
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| BSEP 15-0101 Enclosure 6 Page 11 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description' [TSTF-425_1Brunswick Plant Service Water (PSW) System / Service Water (SW) 3.7.2 3.7.2 System and Ultimate Heat Sink (UHS)
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| Verify cooling tower water level 3.7.2.1 N/A Verify intake structure pump well water level 3.7.2.2 3.7.2.1 Verify water temperature of the UHS 3.7.2.3 3.7.2.2 Operate each cooling tower fan 3.7.2.4 N/A Verify each SW manual, power operated, and automatic valve is 3.7.2.5 3.7.2.3 in the correct position Verify automatic transfer of each DG cooling water supply from N/A 3.7.2.4 the normal SW supply to the alternate SW supply Verify each required SW System automatic actuation 3.7.2.6 3.7.2.5 Diesel Generator (OG) Standby Service Water (SSW) System 3.7.3 N/A Verify each DG SSW System manual, power operated, and 3.7.3.1 N/A automatic valve in the flow path, is in the correct position Verify the DG SSW System pump starts automatically and 3.7.3.2 N/A energizes the respective bus Main Control Room Environmental Control (MCREC) System / 3.7.4 3.7.3 Control Room Emergency Ventilation (CREV) System Operate each MOREC subsystem 3.7.4.1 3.7.3.1 Verify each MCREC subsystem actuates 3.7.4.3 3.7.3.4 Verify each MCREC subsystem can maintain a positive 3.7.4.4 N/A pressure Control Room Air Conditioning (AC) System 3.7.5 3.7.4 Verify each control room AC subsystem can remove the 3.7.5.1 3.7.4.1 assumed heat load Main Condenser Offgas 3.7.6 3.7.5 Verify the gross gamma activity rate of the noble gases 3.7.6.1 3.7.5.1 Main Turbine Bypass System 3.7.7 3.7.6 Verify one complete cycle of each main turbine bypass valve 3.7.7.1 3.7.6.1 Perform a system functional test 3.7.7.2 3.7.6.2 Verify the TURBINE BYPASS SYSTEM RESPONSE TIME is 3.7.7.3 3.7.6.3 within limits Spent Fuel Storage Pool Water Level 3.7.8 3.7.7 Verify the spent fuel storage pool water level 3.7.8.1 3.7.71 AC Sources-Operating 3.8.1 3.8.1 Verify correct breaker alignment and indicated power availability 3.8.1.1, 3.8.1.1 of offsite circuits
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| BSEP 15-0101 Enclosure 6 Page 12 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description' [TSTF-425_ Brunswick Verify each DG starts from standby conditions 3.8.1.2 3.8.1.2 Verify each DG is synchronized and loaded and operates 3.8.1.3 3.8.1.3 Verify each day tank fuel oil 3.8.1.4 3.8.1.4 Check for and remove accumulated water from each day tank 3.8.1.5 3.8.1.5 Verify the fuel oil transfer system operates to transfer oil 3.8.1.6 3.8.1.6 Verify each DG starts from standby condition and achieves and 3.8.1.7 3.8.1.7 maintains voltage and frequency Verify automatic and manual transfer of the unit power supplies 3.8.1.8 3.8.1.8 Verify each DG rejects a load greater than or equal to its 3.8.1.9 3.8.1.9 associated single largest post-accident load Verify each DG does not trip and voltage is maintained during 3.8.1.10 N/A and following a load rejection Verify on an actual or simulated loss of offsite power signal: de- 3.8.1.11 N/A energization of emergency buses, load shedding from emergency buses, and DG auto-starts Verify on an actual or simulated ECCS initiation signal each DG 3.8.1.12 N/A auto-starts from standby condition Verify each DG's automatic trips are bypassed on an actual or 3.8.1.13 3.8.1.10 simulated ECCS initiation signal Verify each DG operates for greater than or equal to 24 hours 3.8.1.14 N/A Verify each DG operating at a power factor greater than or equal N/A 3.8.1.11 0.9 operates for greater than or equal 60 minutes when loaded Verify each DG starts and achieves voltage and frequency 3.8.1.15 N/A Verify each OG: synchronizes with offsite power source, 3.8.1.16 N/A transfers loads to offsite power source, and returns to ready-to-load operation Verify with a DG operating in test mode and connected to its 3.8.1.17 3.8.1.12 bus, an actual or simulated ECCS initiation signal overrides the test mode Veiyinterval between each sequenced load block is within 3.8.1.18 3.8.1.13
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| _+10% of design interval Verify, on an actual or simulated loss of offsite power signal in 3.8.1.19 3.8.1.14 conjunction with an actual or simulated ECCS initiation signal:
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| de-energization of emergency buses, load shedding from emergency buses, and DG auto-starts Verify, when started simultaneously from standby condition, 3.8.1.20 N/A each DG achieves voltage and frequency______
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| Diesel Fuel Oil, Lube Oil, and Starting Air / Diesel Fuel Oil 3.8.3 3,8.3 Verify fuel oil storage tanks inventory 3.8.3.1 3.8.3.1
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| BSEP 15-0101 Enclosure 6 Page 13 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description 1 TSTF-425_[Brunswick Verify lube oil inventory 3.8.3.2 N/A Verify each DG air start receiver pressure 3.8.3.4 N/A Check for and remove water from each fuel oil storage tank 3.8.3.5 3.8.3.3 DC Sources - Operating 3.8.4 3.8.4 Verify battery terminal voltage 3.8.4.1 3.8.4.1 Verify battery terminals and connectors N/A 3.8.4.2 Verify battery cells, cell plates, and racks show no damage N/A 3.8.4.3 Remove visible corrosion and verify anti-corrosion material N/A 3.8.4.4 Verify each required battery charger 3.8.4.2 3.8.4.5 Verify battery capacity is adequate to supply the required 3.8.4.3 3.8.4.6 emergency loads Verify battery capacity is greater than or equal to 80% of the N/A 3.8.4.7 manufacturer's rating Battery Parameters I Battery Cell Parameters 3.8.6 3.8.6 Verify each battery float current 3.8.6.1 N/A Verify each battery pilot cell voltage 3.8.6.2 N/A Verify each battery connected cell electrolyte level 3.8.6.3 N/A Verify each battery cell temperature 3.8.6.4 N/A Verify each battery connected cell voltage 3.8.6.5 N/A Verify battery capacity is greater than or equal to 80% of the 3.8.6.6 N/A manufacturer's rating Verify battery cell parameters meet Table 3.8.6-1 Category A N/A 3.8.6.1 limits Verify battery cell parameters meet Table 3.8.6-1 Category B N/A 3.8.6.2 limits Verify average electrolyte temperature of representative cells N/A 3.8.6.3 Inverters - Operating 3.8.7 N/A Verify correct inverter voltage, frequency, and alignment 3.8.7.1 N/A Inverters - Shutdown 3.8.8 N/A Verify correct inverter voltage, frequency, and alignment 3.8.8.1 N/A Distribution Systems - Operating 3.8.9 3.8.7 Verify correct breaker alignments and indicated power 3.8.9.1 3.8.7.1 availability to required AC and DC electrical power distribution subsystems
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| BSEP 15-0101 Enclosure 6 Page 14 of 16 Technical Specification Cross Reference for TSTF-425 and BrunswickMark-ups Technical Specification Section Title/Surveillance Description 1 JTSTF-425 Brunswick Verify no combination of more than two power conversion N/A 3.8.7.2 modules (consisting of either two lighting inverters or one lighting inverter and one plant uninterruptible power supply unit) are aligned to Division II bus B.
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| Distribution Systems - Shutdown 3.8.10 3,8.8 Verify correct breaker alignments and indicated power 3.8.10.1 3.8.8.1 availability to required AC and DC electrical power distribution subsystems Refueling Equipment Interlocks 3.9.1 3.9.1 Perform CHANNEL FUNCTIONAL TEST on required refueling 3.9.1.1 3.9.1.1 equipment interlock inputs Refuel Position One-Rod-Out Interlock 3.9.2 3.9.2 Verify reactor mode switch locked in Refuel position 3.9.2.1 3.9.2.1 Perform CHANNEL FUNCTIONAL TEST 3.9.2.2 3.9.2.2 Control Rod Position 3.9.3 3.9.3 Verify all control rods are fully inserted 3.9.3.1 3.9.3.1 Control Rod Operability - Refueling 3.9.5 3.9.5 Insert Each withdrawn control rod at least one notch 3.9.5.1 3.9.5.1 Verify each withdrawn control rod scram accumulator pressure 3.9.5.2 3.9.5.2 Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel 3.9.6 3.9.6 Verify RPV water level 3.9.6.1 3.9.6.1 Reactor Pressure Vessel (RPV) Water Level - New Fuel or 3.9.7 3.9.6 Control Rods Verify RPV water level 3.9.7.1 3.9.6.1 Residual Heat Removal (RHR) - High Water Level 3.9.8 3.9.7 Verify one RHR shutdown cooling subsystem is operating 3.9.8.1 3.9.7.1 Residual Heat Removal (RHR) - Low Water Level 3.9.9 3.9.8 Verify one RHR shutdown cooling subsystem is operating 3.9.9.1 3.9.8.1 Reactor Mode Switch Interlock Testing 3.10.2 3.10.2 Verify all control rods are fully inserted in core cells containing 3.10.2.1 3.10.2.1 one or more fuel assemblies Verify no CORE ALTERATIONS are in progress 3.10.2.2 3.10.2.2 Single Control Rod Withdrawal - Hot Shutdown 3.10.3 3.10.3 Verify all control rods, other than the control rod being 3.10.3.2 3.10.3.2 withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed
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| BSEP 15-0101 Enclosure 6 Page 15 of 16 Technical Specification Cross Reference for TSTF-425 and Brunswick Mark-ups Technical Specification Section Title/Surveillance Description 1 TSTF-425 Brunswick Verify all control rods, other than the control rod being 3.10.3.3 3.10.3.3 withdrawn, are fully inserted Single Control Rod Withdrawal - Cold Shutdown_ 3.10.4 3.10.4 Verify all control rods, other than the control rod being 3.10.4.2 3.10.4.2 withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed Verify all control rods, other than the control rod being 3.10.4.3 3.10.4.3 withdrawn, are fully inserted Verify a control rod withdrawal block is inserted 3.10.4.4 3.10.4.4
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| ,Single Control Rod Drive (CRD) Removal,- Refueling, 3.10.5 ,3,1:0.5 Verify all control rods, other than the control rod withdrawn for 3.10.5.1 3.10.5.1 the removal of the associated CRD, are fully inserted Verify all control rods, other than the control rod withdrawn for 3.10.5.2 3.10.5.2 the removal of the associated CRD, in a five by five array centered on the control rod withdrawn for the removal of the associated CRD, are disarmed Verify a control rod withdrawal block is inserted 3.10.5.3 3.10.5.3 Verify no other CORE ALTERATIONS are in progress 3.10.5.5 3.10.5.5 Multiple Control Rod Withdrawal - Refueling ... .. 3.10.6 3.310.6 Verify the four fuel assemblies are removed from core cells 3.10.6.1 3.10.6.1 associated with each control rod or CRD removed Verify all other control rods in core cells containing one or more 3.10.6.2 3.10.6.2 fuel assemblies are fully inserted Verify fuel assemblies being loaded are in compliance with an 3.10.6.3 3.10.6.3 approved reload sequence SHUTOWNMARGN(SM) est-Refelin ........ .!08* 3.10.8 Verify no other CORE ALTERATIONS are in progress 3.10.8.4 3.10.8.4 Verify CRD charging water header pressure 3.10.8.6 3.10.8.6 ReciculaionLoop-Tetin , -' 3.0.9 N/A '
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| Verify LCO 3.4.1 requirements suspended 3.10.9.1 N/A Verify THERMAL POWER during PHYSICS TESTS 3.10.9.2 N/A Training Startups... 3.10.10 .N/A Verify all OPERABLE IRM channels 3.10.10.1 N/A Verify average reactor coolant temperature 3.10.10.2 N/A Notes:
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| : 1. The information provided in the Technical Specification Section Title/Surveillance Description portion of this Enclosure represent summary descriptions of the referenced
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| BSEP 15-0101 Enclosure 6 Page 16 of 16 TSTF 425/BSEP TS Surveillances, provided for information purposes only, and not intended to be verbatim descriptions.
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| : 2. This BSEP Surveillance Frequency is provided in the BSEP Inservice Testing Program.
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| This BSEP Surveillance Frequency is not proposed for inclusion in the Surveillance Frequency Control Program.
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| BSEP 15-0101 Enclosure 7 Page 1 of 2 Proposed No Significant Hazards Consideration Determination
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| ==Subject:==
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| Application For Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements.
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| to a Licensee-Controlled Program Description of Amendment Request This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for General Electric Plants, BWRI4 (NUREG-1 433), to allow relocation of specific technical specification (TS) surveillance frequencies to a licensee-controlled program.
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| The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642), "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b" and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).
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| The proposed changes are consistent with NRC-approved Industry/l-STF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b." The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). This change is applicable to licensees using probabilistic risk guidelines contained in NRC approved Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative Sb, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. MLO71 360456).
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| Basis for the Proposed No Significant Hazards Consideration As required by 10 CFR 50.91 (a), the Duke Energy analysis of the issue of no significant hazards consideration is presented below:
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| : 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
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| Response: No The proposed change relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.
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| 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.
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| The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.
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| Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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| BSEP 15-0101 Enclosure 7 Page 2 of 2
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| : 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
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| Response: No No new or different accidents result from utilizing the proposed change. 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. 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.
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| Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
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| : 3. Does the proposed change involve 'a significant reduction in a margin of safety?
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| Response: No The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Duke Energy will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Revision 1, in accordance with the TS SFCP. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.
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| Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
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| Based upon the reasoning presented above, Duke Energy concludes that the requested change does not involve a significant hazards consideration as set forth in 10 CER 50.92(c), Issuance of Amendment.
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| .- ENERGY DUKE Brunswick Vice President William R. Gideon Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 December 21, 2015 Serial: BSEP 15-01 01 TSC-201 5-03 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
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| ==Subject:==
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| Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Application For Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program Ladies and Gentlemen:
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| In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR Part 50.90), "Application for Amendment of License, Construction Permit, or Early Site Permit,"
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| Duke Energy Progress, Inc. (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BS EP), Unit Nos. 1 and 2. The proposed amendment would modify TS 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 Sb, Risk-Informed Method for Control of Surveillance Frequencies." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.5, "Programs and Manuals." The changes are consistent with Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) Standard Technical Specifications (STS) Change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative Sb," Revision 3 (ADAMS Accession No. ML090850642).
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| Enclosure 1 provides a description of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Enclosure 2 provides documentation of Probabilistic Risk Assessment (PRA) technical adequacy. Enclosure 3 and 4 provide the existing TS pages marked up to show the proposed changes for Units 1 and 2, respectively.
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| Enclosure 5 provides, for information only, the existing Unit 1 TS Bases pages marked up to show the proposed changes. Enclosure 6 provides a cross-reference between the TSTF-425 marked up TS pages and the BSEP Unit 1 and 2 TS pages. Enclosure 7 provides the No Significant Hazards Consideration.
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| Duke Energy requests approval of the proposed license amendment by December 21, 2016, with the amendment being implemented within 120 days.
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| In accordance with 10 CFR 50.91* Duke Energy is providing a copy of the proposed license amendment to the designated representative for the State of North Carolina.
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| U.S. Nuclear Regulatory Commission Page 2 of 3 Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.
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| I declare, under penalty of perjury, that the foregoing is true and correct. Executed on December 21, 2015.
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| William R. Gideon MAT/mat
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| ==Enclosures:==
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| : 1. Evaluation of Proposed License Amendment Request
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| : 2. Documentation of Probabilistic Risk Assessment (PRA) Technical Adequacy
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| : 3. Marked-up Technical Specification and Operating License Pages - Unit 1
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| : 4. Marked-up Technical Specification and Operating License Pages - Unit 2
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| : 5. Marked-up Technical Specification Bases Pages - Unit 1 (For Information Only)
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| : 6. Cross-Reference between TSTF-425, Revision 3 and BSEP Technical Specifications
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| : 7. Proposed No Significant Hazards Consideration Determination
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| U.S. Nuclear Regulatory Commission Page 3 of 3 cc (with enclosures):
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| U.S. Nuclear Regulatory Commission, Region II ATTN: Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U.S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWEN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission ATTN: Ms. Michelle Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)
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| Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1.645 lee.cox@ dhhs.nc.gov
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| BSEP 15-01 01 Enclosure 1 Page 1 of 4 Evaluation of Proposed License Amendment Request
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| ==Subject:==
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| Application For Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program 1.0 Description The proposed amendments would modify the Brunswick Steam Electric Plant (BSEP), Units Nos. 1 and 2, 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 5." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.5, "Programs and Manuals."
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| The changes are consistent with Nuclear Regulatory Commission (NRC) approved Industry/TISTF Standard Technical Specifications (STS) change TSTF-425, Revision 3 (ADAMS Accession No. ML090850642). The availability of the TS improvement was published in the FederalRegister on July 6, 2009 (74 FR 31996).
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| 2.0 Assessment 2.1 Applicability of Published Safety Evaluation Duke Energy Progress, Inc., (Duke Energy) has reviewed the safety evaluation dated July 6, 2009 (74 FR 31996). This review included a review of the NRC staff's evaluation, TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Revision 1 (ADAMS Accession No. ML071360456). includes Duke Energy's documentation with regard to Probabilistic Risk Assessment (PRA) technical adequacy consistent with the requirements of Section 4.2 of Regulatory Guide 1.200, Revision 2 (ADAMS Accession No. ML090410014), and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.
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| Duke Energy has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to BSEP and justify this amendment to incorporate the changes to the BSEP TSs.
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| 2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3, but Duke Energy proposes variations or deviations from TSTF-425, as identified below.
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| *Revised (i.e., typed) TS pages are not included in this amendment request given the number of TS pages affected and the straightforward nature of the proposed changes.
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| Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90, "Application for Amendment of License, Construction Permit, or Early
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| BSEP 15-01 01 Enclosure 1 Page 2 of 4 Site Permit," in that the mark-ups fully describe the changes desired. This represents an administrative deviation from the NRC staff's model application dated July 6, 2009 (74 FR 31996), with no impact on the NRC staff's model safety evaluation published in the same Federal Register Notice. As a result of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the NRC staff's model application. This deviation is consistent with many other industry applications adopting TSTF-425 (e.g., NRC Accession Nos.
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| ML14105A042 (i.e., Turkey Point) and ML14259A564 (i.e., Fermi)).
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| *The definition of STAGGERED TEST BASIS is being retained in BSEP TS Definition Section 1.1 because this terminology is used in TS Programs and Manuals Section 5.5.13, "Control Room Envelope Habitability Program," which is not the subject of this amendment request and is not proposed to be changed. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| *NRC letter dated April 14, 2010 (i.e., ML100990099), provides a change to an optional insert (Insert #2) to the existing TS Bases to facilitate adoption of the Traveler. The TSTF 425 TS Bases insert states the following:
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| The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
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| This statement only applies to frequencies that have been changed in accordance with the Surveillance Frequency Control Program (SFCP) and does not apply to frequencies that are relocated but not changed. Consistent with NUREG-1433, Revision 4 (i.e.,
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| ML12104A193), Duke Energy has replaced the TSTF-425 TS Bases Insert #2 with the following:
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| The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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| For Bases sections addressing multiple affected SRs, the TSTF-425 TS Bases Insert #2 has been revised to state the following:
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| The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
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| *Enclosure 6 provides a cross-reference between the NUREG-1433 Surveillance Requirements (SRs) included in TSTF-425 versus BSEP TS. This Enclosure includes a summary description of the referenced TSTF-425/BSEP TS SRs which is being provided for information purposes only and is not intended to be a verbatim description of the TS SRs. This cross-reference highlights the following:
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| o SRs included in TSTF-425 and corresponding BSEP SRs with identical SR numbers; o SRs included in TSTF-425 and corresponding BSEP SRs with differing SR numbers; o SRs included in TSTF-425 that are not contained in the BSEP TS; and o BSEP plant-specific SRs that are not contained in the TSTF- 425 mark-ups
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| BSEP 15-0101 Enclosure 1 Page 3 of 4 Concerning the above, BSEP SRs that have SR numbers identical to the corresponding TSTF-425 SRs are not deviations from TSTF-425. BSEP SRs with SR numbers that differ from the corresponding TSTF-425 SRs are administrative deviations from TSTF 425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| For TSTF-425 SRs that are not contained in the BSEP TS, the corresponding mark-ups included in TSTF-425 for these SRs are not applicable to BSEP. This is an administrative deviation from TSTF-425 with no impact on the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996).
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| For BSEP plant-specific SRs that are not contained in the mark-ups provided in TSTF-425, Duke Energy has determined that the relocation of the frequencies for these BSEP plant specific SRs is consistent with the intent of TSTF-425, Revision 3, and with the NRC's model safety evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, "Introduction," of the model safety evaluation.
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| The subject plant-specific SRs involve fixed periodic frequencies. In accordance with TSTF-425, changes to the frequencies for these SRs would be controlled under the Surveillance Frequency Control Program. The Surveillance Frequency Control Program provides the necessary administrative controls to require that SRs related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Changes to frequencies in the Surveillance Frequency Control Program would be evaluated using the NRC approved methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1.
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| *SR 3.1.2.1 is included within the scope of this submittal but was not included in TSTF-425, Revision 3. The frequency of SR 3.1.2.1 is encompassed by the intent of TSTF-425 and, therefore, is within the scope of the NRC Model Safety Evaluation (i.e.,
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| ML091800157). The NUREG-1 433 markups within TSTF-425 include a similar core exposure based SR frequency (i.e., SR 3.3.1.1.6). During the NRC review of TSTF-425, Revision 1, an Request for Additional Information (RAI) response (i.e., ML080280272) from the TSTF specifically identified frequencies based on core exposure to be within the scope of TSTF-425 and NEI 04-10. In addition, on July 14, 2015, the NRC approved a similar SR frequency relocation for the Fermi TSTF-425 License Amendment (i.e.,
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| ML15155B416).
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| 3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Determination Duke Energy has reviewed the proposed no significant hazards consideration (NSHC) determination published in the Federal Register on July 6, 2009 (74 FR 31996) and has concluded that the proposed NSHC presented in the Federal Register notice is applicable to BSEP. As such, the NSHC determination for this amendment request is provided as to this license amendment request which satisfies the requirements of 10 CFR 50.91 (a).
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| BSEP 15-0101 Enclosure 1 Page 4 of 4 3.2 Commitments There are no new regulatory commitments contained in this submittal.
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| 3.3 Applicable Regulatory Requirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3, and the NRC's model safety evaluation published in the Notice of Availability dated July 6, 2009 (74 FR 31996). Duke Energy has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the Federal Register notice is applicable to BSEP.
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| 3.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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| 4.0 Environmental Consideration Duke Energy has reviewed the environmental consideration included in the NRC staff's model safety evaluation published in the Federal Register on July 6, 2009 (74 FR 31996). Duke Energy has concluded that the staff's findings presented therein are applicable to BSEP, and the determination is hereby incorporated by reference for this application.
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| BSEP 15-0101 Enclosure 2 Documentation of Probabilistic Risk Assessment (PRA)
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| Technical Adequacy
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| BSEP 15-0101 Enclosure 2 Page 1 of 110 Documentation of Probabilistic Risk Assessment (PRA) Technical Adequacy Table of Contents 1.0 Overview ........................................................................................... 2 2.0 Basis to Conclude that the PRA Model Represents the As-Built, As-Operated Plant.............................. .................................................................. 2 2.1 BSEP PRA Model History...................................................................... 3 3.0 Identification of Permanent Plant Changes Not Incorporated in the PRA Model ............. 4 4.0 Conformance With ASME/ANS PRA Standard.................................................. 4 4.1 Internal Events and Internal Flooding PRA ................................................... 4 4.2 Fire PRA......................................................................................... 5 4.3 External Events and Shutdown Risk .......................................................... 6 4.3.1 High Winds and External Flooding ..................................................... 6 4.3.2 Seismic ................................................................................... 6 4.3.3 Transportation and Nearby Facility Accidents ......................................... 6 4.3.4 Shutdown Risk........................................................................... 7 4.3.5 Conclusions on External Events and Shutdown Risk ................................. 7 5.0 Methodology to be Used to Assess Surveillance Frequency Changes........................ 7 6.0 Key Assumptions and Approximations................... ......................................... 8 6.1 DC Power Availability and Battery Life........................................................ 8 6.2 Loss of Off-Site Power (LOOP) Frequencies ................................................. 8 6.3 Fire Modeling ................................................................................... 9 7.0 Conclusions on PRA Technical Adequacy....................................................... 9 8.0 References ................. ....................................................................... 9 List of Tables Table 1 - BSEP Peer Reviews - All Open Findings & Observations ............................ 11 Table 2 - BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions....................................................................................14 Table 3 - BSEP Fire Peer Review Findings & Observations Resolutions ...................... 44 Table 4 - BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions .................................................................................. 105
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| BSEP 15-0101 Enclosure 2 Page 2 of 110 1.0 Overview Brunswick Steam Electric Plant, Unit No 1. and No. 2. (BSEP) will follow the methodology provided in NEI 04-10, Revision 1 (i.e., Reference 1), to develop a risk informed Surveillance Frequency Control Program (SECP) for control of Technical Specification surveillance frequencies. NEI 04-10 provides guidance for implementation of a generic Technical Specifications improvement that establishes licensee control of surveillance test frequencies for the majority of Technical Specifications surveillances. Existing specific surveillance frequencies will be removed from Technical Specifications for the affected specifications and placed under licensee control pursuant to this methodology.
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| The NEI 04-10 methodology uses a risk-informed, performance-based approach for establishment of surveillance frequencies and is consistent with the philosophy of NRC Regulatory Guide 1.174 (i.e., Reference 2). Probabilistic Risk Assessment (PRA) methods will be used to determine the risk impact of the revised intervals. PRA technical adequacy has been addressed through NRC Regulatory Guide 1.200, Revision 2 (i.e., Reference 3), which endorses the ASME/ANS PRA, RA-Sa-2009 (i.e., Reference 4). External events and shutdown risk impact will be considered quantitatively or qualitatively as described herein.
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| This enclosure demonstrates the technical adequacy of the BSEP PRA model to be used as the basis for the BSEP SFCP, consistent with the requirements of Section 3.3 and Section 4.2 of Regulatory Guide 1.200, Revision 2 (i.e., Reference 3):
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| * Section 2.0 addresses the need for the PRA model to represent the as-built, as-operated plant,
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| * Section 3.0 discusses permanent plant changes that have an impact on those things modeled in the PRA but have not been incorporated in the baseline PRA model.
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| *Section 4.0 demonstrates that the various technical elements of the BSEP PRA have been performed consistently with the ASME/ANS PRA Standard as endorsed in the appendices of RG 1.200. The peer reviews that have been conducted and the resolution of findings from those reviews are included in Tables 1 - 4. These demonstrate that the pieces of the PRA have been performed in a technically correct manner.
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| * Section 5.0 includes a summary of the methodology that will be used to assess the risk under the SFCP.
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| *Section 6.0 identifies the key assumptions and approximations relevant to the results used in the decision-making process. This section provides assurance that the assumptions and approximations used in development of the PRA are appropriate.
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| 2.0 Basis to Conclude that the PRA Model Represents the As-Built, As-Operated Plant The BSEP PRA Model of Record (MOR) is maintained as a controlled document and is updated on a periodic basis to represent the as-built, as-operated plant. Duke Energy procedures provide the guidance, requirements, and processes for the maintenance, update, and upgrade of the PRA:
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| BSEP 15-0101 Enclosure 2 Page 3 of 110
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| : a. The process includes a review of plant changes, relevant plant procedures, and plant operating data as required, through a chosen freeze date to assess the effect on the PRA model.
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| bThe PRA model and controlling documents are revised as necessary to incorporate those changes determined to impact the model.
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| cThe determination of the extent of model changes includes the following:
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| * Accepted industry PRA practices, ground rules, and assumptions consistent with those employed in the ASME/ANS PRA Standard (i.e., Reference 4),
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| * Current industry practices,
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| * NRC guidance (i.e., References 2-3),
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| * Advances in PRA technology and methodology, and
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| * Changes in external hazard conditions.
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| For plant changes of small or negligible impact, the model changes can be accumulated and a single revision is performed at an interval consistent with major PRA revisions. The results of each evaluation determine the necessity and timing of incorporation of a particular change into the PRA model. An electronic tracking database is utilized to document pending model changes and updates. Previous major PRA model revisions are summarized in Section 2.1.
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| 2.1 BSEP PRA Model History In response to the original Generic Letter 88-20, the 1992 Individual Plant Examination (IPE) was developed to address the risk from internal initiating events including internal flooding. In 1995, this was expanded to create the Individual Plant Examination of External Events (IPEEE) which included seismic events, internal fires, high winds and tornados, external floods, and transportation and nearby facility accidents. The BSEP PRA model has undergone numerous updates and reviews since the original development to maintain a representation of the as-built, as-operated plant in response to improvements in PRA technology and state-of-the-art methodologies. This section presents summaries of the BSEP PRA MOR updates.
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| MOR 2004 addressed Findings and Observations (F&Os) from the 2001 Peer Review along with a failure and unavailability data and success criteria update. MOR 2005 updated the Human Reliability Analysis (HRAs) and answered F&Os from the 2001 Peer Review to support Mitigating System Performance Index (MSPI). MOR 2006 incorporated the ability to cross-tie service air between units. The MOR 2004, 2005, and 2006 revisions did not change methodologies and did not meet the criteria of an upgrade. An independent peer review was performed in 2007 to assess the status of the PRA internal events model compared to the latest PRA Standard and Regulatory Guide 1.200 requirements.
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| MOR 2007 included the implementation of a new diesel room heat-up analysis, the addition of two independent generators to support DC power, and other PRA model improvements consistent with plant design and operation. In 2008, additional minor changes were made and the PRA was re-issued as MOR 2008. MOR 2010 incorporated minor changes to support Regulatory Guide 1.200 requirements and the internal flooding model was upgraded.
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| This PRA model was then subject to a full scope internal events Peer Review conducted by the BWR Owners Group in 2010.
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| MOR 2011 addressed a majority of F&Os from the 2010 Peer Review. Many of the F&Os
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| BSEP 15-01 01 Enciosure 2 Page 4 of 110 were on the recently upgraded Internal Flooding model, for which this effort resolved technical issues and improved the documentation. MOR 2013 incorporated updates to internal events data such as reliability, unavailability, common cause, initiating events, and HRA data. The MOR 2013 scope was limited and as such, it was not considered a PRA upgrade.
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| 3.0 Identification of Permanent Plant Changes Not Incorporated in the PRA Model The current BSEP Model of Record (MOR 2013) is being used for current applications. All permanent plant modifications and ECs that have been implemented since MOR 2013 have been reviewed as part of Duke Energy PRA model maintenance procedures. There are currently no identified permanent plant modification that have a significant impact on the PRA that have not been incorporated into the MOR.
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| 4.0 Conformance With ASME/ANS PRA Standard The following sections describe the conformance and capability of the BSEP PRA against the ASME/ANS PRA Standard (i.e., Reference 4).
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| 4.1 Internal Events and Internal Flooding PRA The following peer reviews have been conducted to ensure the internal events and internal flooding PRA meets the requirements of ASME/ANS PRA Standard:
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| * In 2001, an industry peer review was performed in accordance with Revision A-3 NEI draft "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance" NEI 00-02 dated June 2, 2000.
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| * In 2007, an industry peer review was performed in accordance with Revision A-3 NEI draft "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance", NEI 00-02, dated June 2, 2000. The model was evaluated against ASME-RA-Sb-2005, Addendum B and Regulatory Guide 1.200.
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| * In 2010, an industry peer review was performed in accordance with NEI 05-04 process. The model was evaluated against ASME PRA Standard ASME/ANS RA-Sa-2009 and Regulatory Guide 1.200, Rev 2.
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| Each of the peer reviews evaluated the model for the supporting requirements for internal events, internal flooding, and containment performance (i.e., Large Early Release Frequency).
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| A total of five findings from the latest 2010 peer review that have not yet been resolved are listed in Table 1. All other findings have been reviewed and resolved and can be found in Table 2. These tables provide a complete list of all the findings, resolutions, and dispositions to determine whether the finding has any significant impact on the 5b application.
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| In reviewing the BSEP risk informed License Amendment Request (LAR) for implementation of NFPA 805, the NRC staff evaluated the quality of the internal events PRA model used to support development of the Fire PRA. The objective of the quality review was, "to determine whether the plant-specific PRA used in evaluating the proposed LAR is of sufficient scope, level of detail, and technical adequacy for the application." The results of the NRC staff quality review are documented in the BSEP NFPA 805 Safety Evaluation for transition to a risk-informed, performance-based fire protection program, ADAMS Accession Numbers ML14310A808 (i.e., Reference 5). The staff concluded that the licensee has demonstrated
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| BSEP 15-0101 Enclosure 2 Page 5 of 110 that the internal events PRA model is technically adequate to support the NEPA 805 risk calculation necessary for the license amendment. While this evaluation was not specific to the 5b application, it further demonstrates the technical adequacy of the internal events model.
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| Based on results of the peer reviews and resolutions, the BSEP internal events and internal flooding PRA meets the requirements of the ASME/ANS PRA Standard as clarified by Regulatory Guide 1.200, Revision 2, at an appropriate capability category to support the BSEP Surveillance Frequency Control Program (SECP). The internal events and internal flooding PRA will be used in accordance with NEI 04-10 to assess proposed surveillance frequency changes under the SFCP.
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| 4.2 Fire PRA The BSEP fire PRA was developed using the guidance provided by NUREG/CR-6850 (i.e.,
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| Reference 7) in support of NFPA 805 fire protection program. The fire PRA is built upon the internal events PRA which was modified to capture the effects of fire. In 2012, an industry peer review was performed on the FPRA using NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Guidelines (i.e., Reference 8). The model was evaluated against the ASME PRA Standard ASME/ANS RA-Sa-2009 (i.e., Reference 4), as clarified by Regulatory Guide 1.200 Rev. 2 (i.e., Reference 3). Of the 263 Supporting Requirements that were determined to be applicable, 208 were assessed as meeting Capability Category II or better, 18 were assessed as meeting Capability Category I, and 36 were assessed as Not Met. The peer team also identified one Unreviewed Analysis Method (UAM). Findings were issued for the UAM and for any Supporting Requirement assessed as not meeting at least Capability Category I1. All findings have been addressed either through model changes, additional documentation, or better justification of the basis. Table 3 provides a complete list of all the findings, resolutions, and dispositions to determine whether the finding has any significant impact on the 5b application.
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| The findings and associated resolutions were submitted as part of the NFPA 805 LAR. The results of the NRC staff quality review of the Fire PRA are documented in the BSEP NFPA 805 Safety Evaluation for transition to a risk-informed, performance-based fire protection program, ADAMS Accession Number ML14310A808 (i.e., Reference 5). The quality review concluded that the technical adequacy and quality of the BSEP PRA is sufficient, with the implementation of certain changes (i.e., Implementation Item #13 in Table S-2 of the NFPA 805 LAR), to support risk-informed changes to the NFPA 805 fire protection program. Those implementation items have been incorporated into the Fire PRA, and a focused scope peer review has been conducted, where required.
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| Between September 2014 and May 2015, a focused scope peer review was performed for the incorporation of sensitive electronics in the FPRA and for the Appendix L sensitivity analysis, pursuant to Implementation Items #13.2 and 15, respectively. The peer review was conducted using the general process defined in NEI 07-12, Revision 1, and focused on FPRA technical elements FSS, IGN, FQ, and UNC in ASME/ANS RA-SA-2009, as endorsed by RG 1.200, Revision 2. Within these constraints, the following Supporting Requirements were assessed as not applicable: FSS-A3, FSS-B1, FSS-B2, FSS-C7, FSS-C8, FSS-D9, FSS-E1, FSS-E2, FSS-F1, FSS-F2, FSS-F3, FSS-G1, FSS-G2, FSS-G3, FSS-G4, FSS-G5, FSS-G6, IGN-A2, IGN-A3, IGN-A4, IGN-A6, and FQ-F2. As documented in the peer review report (i.e.,
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| Reference 12), the remaining Supporting Requirements were assessed as meeting Capability Category II or better, with no Findings.
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| BSEP 15-01 01 Enclosure 2 Page 6 of 110 Based on the results of these peer reviews and resolution, the BSEP fire PRA meets the requirements of the ASME/ANS PRA Standard as clarified by Regulatory Guide 1.200, Revision 2, at an appropriate capability category to support the BSEP SFCP. The fire PRA will be used in accordance with NEI 04-10 to assess proposed surveillance frequency changes under the S FOP.
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| 4.3 External Events and Shutdown Risk The following sections describe how external events and shutdown risk are evaluated.
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| 4.3.1 High Winds and External Flooding BSEP developed both the high winds and external flooding PRA models in 2011. In 2012, an industry peer review was performed which evaluated the models against the ASME PRA Standard ASME/ANS RA-Sa-2009 (Reference 4), as clarified by Regulatory Guide 1.200 Revision 2 (i.e., Reference 3). This was a full-scope review of all the Technical Elements of the high wind and external flood events, at-power PRA. All of the external flooding findings from the 2012 Peer Review have been reviewed and resolved and can be found in Table 4.
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| The high winds have been addressed and will be incorporated for use in this application. The resolutions can also be found in Table 4.
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| Based on results of the peer review and resolutions, the BSEP high winds and external flooding PRA meets the requirements of the ASME/ANS PRA Standard as clarified by Regulatory Guide 1.200, Revision 2, at an appropriate capability category to support the BSEP SFCP. The high winds and external flooding PRA will be used in accordance with NEI 04-10 to assess proposed surveillance frequency changes under the SFCP.
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| 4.3.2 Seismic For the IPEEE submitted in 1995, BSEP employed EPRI's Seismic Margins Analysis (0.3g Review Level Earthquake) to identify vulnerabilities to seismic events. In 2014, BSEP completed a Seismic Hazard Evaluation and Screening Report in response to NRC recommendations of the Near-Term Task Force (NTTF) review of insights from the Fukushima Dai-ichi accident (i.e., Reference 11). The results of the review have shown that the IPEEE is adequate to support screening of the updated seismic hazard for BSEP and that the risk insights obtained from the IPEEE are still valid under the current plant configuration. SSCs impacted by frequency changes under the SFCP, therefore, will be assessed against the seismic margins analysis and evaluated in accordance with NEI 04-10.
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| 4.3.3 Transportation and Nearby Facility Accidents For the IPEEE submitted in 1995, BSEP concluded that potential accidents associated with nearby air traffic, runways, roads, railways, waterways, pipelines, and fixed military and industrial facilities are not considered a significant hazard. Structures, systems, and components (SSCs) impacted by frequency changes under the SFCP, therefore, will be assessed against the IPEEE analysis and evaluated in accordance with NEI 04-10.
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| BSEP 15-0101 Enclosure 2 Page 7 of 110 4.3.4 Shutdown Risk BSEP operates under a shutdown risk management program to support implementation of NUMARC 91-06 (i.e., Reference 10). The shutdown risk management implementing procedure provide guidelines for outage risk management which focuses on proper planning, conservative decision making, maintaining defense in depth, and controlling key safety functions. BSEP will use the shutdown risk management program procedures to assess shutdown risk for proposed surveillance frequency changes.
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| 4.3.5 Conclusions on External Events and Shutdown Risk External hazards screenings have been performed for BSEP to support requirements of the IPEEE and in review of insights from the Fukushima Dai-ichi accident. NEI 04-10 allows for proposed surveillance frequency change evaluations to use hazard screening in the absence of external hazards PRA models. In cases where these methodologies are not appropriate for a surveillance frequency change evaluation, other qualitative or bounding analysis will be utilized to provide justification for the acceptability of the proposed surveillance frequency change. BSEP will follow the NEI 04-10 guidance to assess external event and shutdown risk associated with potential surveillance frequency changes.
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| 5.0 Methodology to be Used to Assess Surveillance Frequency Changes Existing Duke Energy procedures derived from the NEI 04-10 guidance will be used to govern the SFCP and the surveillance test interval (STI) evaluation process. The following steps will be used to assess proposed changes within the BSEP program.
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| *Each STI revision will be reviewed to determine whether there are any commitments made to the NRC that may prohibit changing the interval. If there are no related commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision will proceed. If a commitment exists and the commitment change process does not permit the change, then the STI revision will not be implemented. Only after receiving formal NRC approval to change the commitment will an STI revision proceed.
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| *Systems engineering evaluations and quantitative assessments from available PRA models will be developed for each proposed change. The BSEP internal events, internal flooding, high winds, external flooding, and fire PRAs all meet the requirements for Regulatory Guide 1.200, Revision 2 at appropriate capability categories, and will be used to assess whether an SSC is affected by the proposed STI change. In calculating SSC failure rates, if the breakdown between the standby time-dependent failure rate and the demand-related failure rate probability for affected SSCs is unknown, then the total failure probability will be assumed to be time-related to obtain the maximum test-limited risk condition. The total and cumulative effects on Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) will be assessed, and cumulative risk will be tracked.
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| *If an SSC being assessed is not modeled in the PRA, then appropriate qualitative or bounding risk analysis will be performed for that SSC. Duke Energy procedures derived from NEI guidelines will be used to determine if the qualitative analysis is sufficient for consideration.
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| BSEP 15-0101 Enclosure 2 Page B of 110
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| *Hazard screening performed for the IPEEE, and review of programmatic assessments performed in response to the Fukushima Dai-ichi accident will be used to assess seismic and transportation and nearby facilities accidents for potential changes in STI. The BSEP shutdown risk management program for implementation of NUMARC 91-06 will be used to assess the shutdown risk.
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| *The results of each STI assessment will be documented and presented to an Expert Panel, referred to as the Integrated Decision-making Panel (lDP). The lOP will normally be the same panel used for Maintenance Rule implementation but with the addition of specialists with experience in surveillance testing and system or component reliability. If the IDP approves the STI revision, the change will be documented and implemented, and will be available for audit by the NRC. If the IDP does not approve the STI revision, the surveillance frequency is left unchanged.
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| *Performance monitoring will be conducted as recommended by the IDP. In some cases, no additional monitoring may be necessary beyond that already conducted under the Maintenance Rule. Performance monitoring helps to confirm that no failure mechanisms related to the revised STI become important enough to alter the information provided for the justification of the interval changes.
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| *The lDP is responsible for periodic review of performance monitoring results. If it is determined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the IDP will reset the STI to the previously acceptable test interval.
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| 6.0 Key Assumptions and Approximations A list of potential contributors to the uncertainty in the PRA was compiled. The list below represents the modeling assumptions and uncertainty that are considered to have the greatest impact on the BSEP PRA results if different reasonable alternative assumptions were utilized. The approaches taken for the assumptions below represent industry best practices and therefore the need for sensitivity analyses will be determined separately for each of the individual surveillance frequency changes evaluated.
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| 6.1 DC Power Availability and Battery Life The DC power system at BSEP is one of the largest contributors to plant risk. Determination of battery depletion times and associated accident sequence timing and related success criteria can potentially have an impact on results.
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| 6.2 Loss of Off-Site Power (LOOP) Frequencies Loss of off-site power initiating events have been shown to be important contributors to plant core damage due to the potential for station blackout and the reliance of many frontline systems on AC power. The LOOP initiator was separated into plant, grid, switchyard and weather induced LOOPs, which allowed the model to apply recovery actions to the higher frequency events (i.e., plant and switchyard). BSEP used generic industry data to calculate
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| BSEP 15-0101 Enclosure 2 Page 9 of 110 LOOP frequencies. The LOOP frequency has an impact on ODE and Emergency Diesel Generator (EDG) importance.
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| 6.3 Fire Modeling Fire modeling, although following the technical guidance of NUREG/CR-6850, contains several risk important elements that are judged to contain uncertainties for their respective elements of fire risk methodology. These elements include the fire ignition frequency, heat release rates, fire growth curves, fire suppression failure probabilities, severity factors, and Post-initiator human failure event probabilities. While the approaches taken in the BSEP Fire PRA represent the "state of the art" methodology, they are still constrained by the relatively limited data on fire events at Nuclear Power Plants.
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| 7.0 Conclusions on PRA Technical Adequacy The BSEP PRA model is sufficiently robust and suitable for use in risk informed processes such as the Surveillance Frequency Control Program. The peer reviews that have been conducted and the resolution of findings from those reviews demonstrate that the pieces of the PRA have been performed in a technically correct manner. The assumptions and approximations used in development of the PRA have also been reviewed and are appropriate for their application. Duke Energy procedures are in place for controlling and updating the models, when appropriate, and for assuring that the model represents the as-built, as-operated plant. The conclusion, therefore, is that the BSEP PRA model is acceptable to be used as the basis for risk-informed applications including Risk-Informed Technical Specifications (RITS) Initiative 5b.
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| 8.0 References
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| : 1. NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1, April 2007.
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| : 2. Regulatory Guide 1.174, "An Approach for Using ProbabilisticRisk Assessment in Risk-In formed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, U.S.
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| Nuclear Regulatory Commission, March 2011.
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| : 3. Regulatory Guide 1.200, "An Approach for Determiningthe Technical Adequacy of ProbabilisticRisk Assessment Results for Risk- Informed Activities," Revision 2, U.S.
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| Nuclear Regulatory Commission, March 2009.
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| : 4. ASME/ANS RA-Sa-2009, "Standardfor Level 1/Large Early Release Frequency ProbabilisticRisk Assessment for Nuclear Power PlantApplications," Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
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| : 5. Brunswick Steam Electric Plant, Units 1 and 2 - Issuance of Amendment regarding Transition to a Risk Informed, Performance-Based Fire Protection Program in Accordance with 10CFR 50.48(0) (TAO NOS. ME9623 and ME9624), January, 28, 2015. (ADAMS Accession No. ML14310A808)
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| : 6. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance,"~
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| Revision A-3 draft, June 2, 2000.
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| : 7. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"
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| Volume 2, September, 2005.
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| BSEP 15-01 01 Enclosure 2 Page 10 of 110
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| : 8. NEI 07-12, Draft Version E, "Fire ProbabilisticRisk Assessment (FPRA) Peer Review Process Guidelines," Nuclear Energy Institute, May 2007.
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| : 9. NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard(Internal Events)," Revision 1 (Draft), Nuclear Energy Institute, November 2007.
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| : 10. NUMARC 91-06, "Guidelines for Industry Actions to Address Shutdown Management,"
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| December 1991.
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| : 11. BSEP 14-0028, "Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident." March 31, 2014. ADAMS Accession No. ML14106A461
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| : 12. EPM Report R2427-0001-01-00, "Focused Scope Peer Review for the Brunswick Fire PRA Against the ASME PRA Standard Requirements," September 2014-May 2015.
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| : 1. BSEP Peer Reviews All Open Findings & Observations________
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| SRIF&O # JFinding S~~~~~Table
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| ]Resolution -
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| 5b Impact In BNP-PSA-034, Attachment 3 Instead of 0.008 for single train, a higher value of 0.01 was There is no describes the Type A human used. For the common cause value of 8E-4, a screening impact to the 5b error screening methodology value of 5E-03 was used. The impact of the higher HEPs is application.
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| employed for the Brunswick negligible with a FV value of 5E-03. The ambiguity in the PRA model. The methodology documentation will be addressed at a later point.
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| follows the screening methodology suggested in NUREG 1792. The screening values used in Tables 1 and 2 contradict the values said to be employed in Section E.3.1 (i.e.
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| 0.008 applied to Type A human errors affecting a single train HR-12 and 0.0008 for common cause (CAT I/Il/Ill) human failure events).
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| 2-3 Nonetheless, Tables 1 & 2 in Attachment 3 utilize screening values of 0.01 for single train Type A human errors and 5E-03 for common cause errors. When these screening values were used for the HEPs and the model subsequently quantified, all cross-train Type A HEPs using screening values of 5E-03 having a Fussel-Vesely Importance < 5E-03 are stated to be screened.
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| Table 1. BSEP Peer Reviews - All Open Findings & Observations SR/F&O # Finding Resolution 5b Impact As discussed in BNP-PSA-049, No changes made to the model. Any credit of repair would Meets CAT I, Appendix D, Section D.1, the reduce LERF. Meets CAT I, which provides conservative which provides CET structure allows for the results for LERF. conservative identification of recovery and results for LERF.
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| repair actions that can terminate There is no or mitigate the progression of a impact to the 5b severe accident. This process application.
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| was incorporated into the LEC3TI original analysis, rather than (CAT I) performing a review of 3-12 significant accident progression sequences and then incorporating repair, as would be inferred from the standard.
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| However, it does not appear that significant accident progression sequences were reviewed.
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| There is no evidence that No changes made to the model. Any credit for equipment Meets CAT I, significant accident sequences survivability would reduce LERF. Meets Cat I, which which provides were reviewed to determine if provides conservative results for LERF. conservative LE-C10 engineering analyses could results for LERF.
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| (CAT I) support continued equipment There is no LE-C12 operation or operator actions to impact to the 5b reduce LERF. It was noted that application.
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| (CAT I) this conservative approach with 3-12 respect to equipment survivability was documented in the uncertainty analysis (BNP-PSA-075, Table 1, Item 236).
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| Table 1. BSEP Peer Reviews - All Open Findings & Observations SR/F&O # Finding Resolution 5b Impact BNP-PSA-049, Section 3.1.2 No changes made to the model. Any additional treatment of Meets CAT I, notes that the treatment of scrubbing would reduce LERF. Meets CAT I, which which provides scrubbing by the reactor provides conservative results for LERF. conservative LE-C13 building is treated in a results for LERF.
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| (CAT I) conservative method. This There is no conservative approach was impact to the 5b 3-13 identified in the uncertainty application.
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| analysis (BNP-PSA-075, Table 1, Item 217).
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| There does not appear to be a Standard EPRI PRA codes were used in the analysis. Documenting centralized discussion of Documentation of computer code limitations will be computer code computer code limitations, addressed at a later point, limitations will not Codes used in generic change the references and SAR (and quantified risk SC-C2 associated limitations) are not metrics. There is (CAT I/ll/Ill) discussed. System level no impact to the success criteria for meeting Top 5b application.
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| 6-8 Event gate success criteria are not as well documented in BNP-PSA-033 but appear to be documented on a distributed basis in Heatup Calculations
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| ___________and in System Notebooks. __________________________ ________
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact Unavailability data was taken EDG unavailability was corrected to include all unavailability Resolution of this from MR databases, and outage hours accrued regardless of whether the plant was in finding corrected periods were excluded from the operation or outage condition. In the unavailability value the EDG data to UA calculations (as documented calculation instead of using the unavailability hours while at include all in the BNPUnavail spreadsheet power and dividing them by the total at power hours the accrued that is part of the BNP-PSA-004 value was determined by dividing all unavailability hours unavailability.
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| notebook). In the case of the accrued over the entire window of data collection. The There is no shared EDGs, the BNPUnavail correction resulted in an as expected increase in EDG impact to the 5b spreadsheet is incorrectly unavailability values. application.
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| deleting all EDG QOSs that DA-C1 3 occur when either unit is in an (CAT i/Ill1) outage. Based on the modeling of the shared EDGs, the correct 1-2 treatment would be to NOT exclude any EDG outage (regardless of unit outage condition). This error results in an EDG unavailability that is too low. It should be noted, that based on current data, the impact of this error appears to not be too significant on the results.
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| For SBO, the event tree model An SBO in the PRA model assumes a loss of suppression Resolution of this does not guarantee that a safe pool cooling due to the total loss of AC power on the finding expanded stable state has been achieved, affected unit. There are LOSP cut sets in the PRA results the event trees to AS-A2 The LOSP convolution analysis where power is available and that suppression pool cooling address the (CAT I/Il/Ill) (BNP-PSA-036) includes failed. The recovery of AC power is a subset of those required recovery AC power implicitly. cutsets so that suppression pool cooling is not required to questions for a 1-15 However, the accident be modeled. safe stable state.
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| sequence analysis does not There is no consider the possibility of failure However, for the SBO event tree, there were some non- impact to the 5b of suppression pool cooling conservatism noted, including the no questions involving application.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact following AC power recovery. suppression pool venting or addition actions the operator must perform to maintain long term RCIC operation. The event tree was expanded to address required questions for a safe stable state. The accident sequence notebook was updated (Mar. 2011) to address the modified event tree.
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| Based on the information OSI P1 data was used to make estimates of the component Resolution of this presented in the system standby time for systems with redundant trains. A review of finding collected notebooks (BNP-PSA-062), the the plant specific operating data over a period from 2003 to plant data to PRA estimates the standby time 2008 generally shows that the system operating load is estimate the for components based on the balanced equally across the available trains. In instances component number of trains available vs. where there are two available trains and only one is required standby time for the number of running trains to operate there is a near 50/50 split balance between the redundant trains.
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| DA-C8 (e.g., 50% for a 1 of 2 system). pair. This is similarly seen in the systems with 3 trains with The plant data (CAT I) Actual plant specific data each train carrying approximately 33% of the operating load, supports the concerning standby time is not Based on the results of the data review, in cases where current values collected and evaluated in the there are standby components they are given a standard used. This finding 22 BSEP PRA. However, may be 50% or 33% running time per component as a data is sufficiently available from 0SI Pl to support supported approximation. resolved for the this. SR to be assessed as meeting CAT IlI/ll.
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| There is no impact to the 5b application.
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| For small break LOCA, the high A specific MAAP analysis was performed which verified Resolution of this SB3 end of water break is RCIC as a success path for a 1-inch diameter break. finding performed (CTIl/l) approximately 1" dia., RCIC is a MAAP analysis.
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| (CTII/l) credited for HPI for success, butThrisn no MAAP run was performed to impact to the 5b 1-1 demonstrate the success. application.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O #__ Finding fResolution [ 5b Impact There is no evidence of Resolved in the BSEP Model; the flooding findings have Resolution of this documentation of review or been resolved by improved documentation and PRA model finding improved modification of accident analysis. documentation sequences. Discussions with and PRA model the PRA Staff indicate that IF analysis. This lEs are evaluated using the finding is Transient ET. Specific flood sufficiently impacted equipment is taken out resolved for the of service in conjunction with the SR to be specific flooding IE. assessed as meeting CAT All sequences were added to I/Il/Ill. There is no the general transient event impact to the 5b sequence. No new event trees application.
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| were generated. The IFQU-A1 quantification section of the (NOT MET) flooding report does not discuss how it was determined that 2-9 there were no special flooding sequences that warranted special handling.
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| The Flooding Analysis document RSC 10-05 does not provide information concerning modeling of system/component failures due to pipe failures in that system, as opposed to failure due to flooding and flood propagation. Detailed discussions were necessary to establish the logic by which, for example, Fire Protection Water and Service Water systems can
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact still be credited for injection to the vessel after piping failures in the respective systems.
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| Sequence modifications are not documented. Some fire protection and service water pipe breaks will fail the system as a LPI source by failing the injection path, including consideration of flow diversion effects; this may not be fully accounted for.
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| Operator interview insights are Detailed operator interviews were conducted for the Resolution of this documented in the HRA purpose of confirming procedure interpretations. PRA finding updated Calculator. The information documents have been updated to improve their clarity in PRA documents contained in the HRA Calculator this area. The HFEs mentioned are no longer used in the to improve clarity.
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| was sufficient to demonstrate PRA. This finding is the Capability Category I was sufficiently met. However, the information in The results of operator interviews are put into the HRA resolved for the the HRA did not demonstrate calculator. If there were any special comments from the SR to be that detailed talk throughs with operators, they are included in the operator response tab assessed as HR-E3I Operations and Training for each operator action. meeting CAT IlIlll.
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| (A ) Personnel were conducted for There is no the purpose of confirming impact to the 5b procedure interpretations. For application.
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| example, many of the calculations referred only to an interview conducted with a single operator on 9116-17/2008. A few calculations referred a "talk through" in January 2008, an operator interview on 3/11/2010, or
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution j 5b Impact simulator runs conducted on 1/19/2010. A few calculations (OPER-BLACKSTART, OPER-CNS, OPER-CWSIE) did not
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| - have any input on operator interviews. The purpose and content of these interviews is not evident. Based on the information provided, Capability i/I/ll was not demonstrated __________________________ ________
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| While it was documented that The actions for which there are observed simulator runs are Resolution of this simulator observations and talk- in the Annex section with applicable documentation in the finding added throughs were performed in main body of the calculation to ensure that it can be traced. documentation of most HRA calculations, there is A generic operator discussion sheet was added to the both simulator no evidence that these calculation (BNP-PSA-034). interviews and observations or talk-throughs checklists. This HE4 were used to confirm the The results of operator interviews are put into the HRA finding is (CAT I) response models for the calculator. If there were any special comments from the sufficiently scenarios modeled in the PRA. operators, they are included in the operator response tab for resolved for the For example, there was no each operator action. SR to be interview checklist, assessed as simulator/scenario checklist, or meeting CAT IlI/ll.
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| other documentation to There is no demonstrate that the HRA impact to the 5b analyst confirmed the response application.
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| models.
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| Problems were noted with the OPER-DCDG has been cleaned up and execution errors Resolution of this (CRT-II HRA calculation for OPER- have been added. Additionally the DCDG modification has finding revised the (CAT-G3 DCDG. Specifically, no been revised to eliminate the need to make the connections execution R-3 execution failure probabilities discussed. The HRA has been revised to reflect the change evaluation. There (CTI/l) were assigned to the tasks of in DCDG operation. is no impact to the 3-8 starting and connecting the DG. 5b application.
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| ____________Additionally, the calculation may
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact not have considered all of the necessary breaker manipulations.
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| In general, the HRA calculator All the HFE's have been gone through and information such Resolution of this file was reviewed and found to as training has been added if applicable. If a HEP has no finding added the provide an assessment of the value, (no cognitive procedure or training is not available) missing performance shaping factors then it is documented in the corresponding tab. information to the listed in the SR for the HEP HRA calculator.
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| calculations. Some detail in the There is no calculations could be enhanced. impact to the 5b For example, the operator application.
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| action OPER-LDSHD calculation does not have the HR-G3 cognitive procedure listed and (CAT IlIll1) does not address the training requirements. Calculations for 3-6 OPERMSIVCBP and OPER-DEPRESS1 state that simulator and classroom training are provided but does not provide a frequency. The calculations for OPER-DCDG and OPER-N2SUPPLY do not address training, the cognitive procedure or the staffing requirements.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The only review for initiating A systematic evaluation was performed and documented in Resolution of this events caused by systems the IE Notebook, Calculation BNP-PSA-032, Rev. 9, Section finding performed failure was for mitigating 2.4. Section 2.4 contains a table in which all of the BSEP a systematic systems already modeled. No plant systems are listed and reviewed for their ability to evaluation of each evidence of reviews of other result in a plant trip/initiating event. system as a IE-A5 systems was found. potential IE (NOT MET) contributor. This finding is 4-2 sufficiently resolved for this SR to be assessed as meeting CAT II.
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| There is no impact to the 5b application.
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| The use of a single pipe section The pipe rupture frequencies for Break Outside Resolution of this as the basis for IE frequency for Containment have been revised to incorporate the latest finding breaks outside of containment Pipe Rupture Frequencies from EPRI's Pipe Rupture incorporated the could be nonconservative by a Frequencies for Internal Flooding Probabilistic Risk updated EPRI factor of 100 to 1000 compared Assessments, Revision 2 Report 1021086. Calculation pipe failure to latest EPRI pipe failure BNP-PSA-032, Revision 9 Section 3.2 contains the analysis frequencies into frequency methodology (see using the current methods of calculating pipe break the internal IE-C1 also IE-C6). frequencies by linear foot of piping. The analysis assesses flooding model.
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| (CAT I/Il/Ill) piping systems that can contribute to high energy line There is no breaks outside containment and includes Main Steam Lines, impact to the 5b 1-3 Feedwater Lines, High Pressure Coolant Injection Lines, application.
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| Reactor Core Isolation Cooling Lines, Reactor Water Clean Up Lines and Scram Discharge Volume.
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| The analysis identified the need to include Main Steam Lines breaks as an initiator and Nuclear Task Management
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| #00458348 was generated to include this initiator in future model revisions. The increase in IE frequency of the main _________
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact steam lines was a result of identifying and including additional piping and a common steam chest.
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| The other large BOC initiators analyzed with the new method resulted in un-isolated pipe break frequencies that were near the values of the old EPRI pipe break analysis values.
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| The PRA uses a 82% plant Revised the plant capacity factor to exclude the period from Resolution of this
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| -capacity factor. (Section 4.1.2 of 1989 through 1993. The resulting capacity factor of 92.4% finding updated BNP-PSA-032) This appears is based on a 15 year period from 1994 through 2008 and is the plant capacity nonconservative based on the considered more representative of current and future factor to a more IE-C5 past fifteen years. Capacity operations (Section 4.1.2 of BNP-PSA-032 Rev 9). representative factor appears to be over 90% value. There is (CTII) over past 15 years and that no impact to the
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| -1 would be the expectation for the 5b application.
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| future. Use of a 82% capacity factor results in underestimating IE frequencies on a reactor critical year basis.
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| There are questionable The pipe rupture frequencies for Break Outside Resolution of this assumptions on Breaks Outside Containment have been revised to incorporate the latest finding reanalyzed Containment, particularly the Pipe Rupture Frequencies from EPRI's Pipe Rupture pipe breaks use of a nonconservatively Frequencies for Internal Flooding Probabilistic Risk outside EC6 small pipe break frequency for Assessments, Revision 2 Report 1021086. Calculation containment with (OME) Main Steam and Main BNP-PSA-032, Revision 9 Section 3.2 contains the analysis the updated EPRI (OME) Feedwater piping through use of using the current methods of calculating pipe break methodology.
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| 64 old EPRI pipe section frequencies by linear foot of piping. The analysis assesses The analysis did methodology vice consideration piping systems that can contribute to high energy line not change the
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| -of piping length. breaks outside containment and includes Main Steam Lines, prior screening Feedwater Lines, High Pressure Coolant Injection Lines, results.
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| Reactor Core Isolation Cooling Lines, Reactor Water Clean Documentation Up Lines and Scram Discharge Volume. The analysis, was added and
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact using the new method resulted in un-isolated pipe break this finding is frequencies that were near the values of the old EPRI pipe considered break analysis values and thus the high energy Break sufficiently Outside Containment initiators are still screened from the resolved for the model. SR to be assessed as meeting CAT 1/11/111. There is no impact to the 5b application.
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| An older EPRI pipe break The pipe break frequency methodology has been update to Resolution of this method based upon piping the EPRI TR-1013141 method. See Table F.15 (RSC finding updated segments is used. This 05) for break frequency categories and Table F.16 (RSC the pipe break methodology had been the 05) for pipe break frequencies. frequency subject of previous F&O IF-D5a- methodology.
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| : 1. The latest EPRI methodology This finding is is based upon piping length, and sufficiently differentiates between whether resolved for the the Service Water is sea water, SR to be lake water, or river water. The assessed as lFEEV-A5 Brunswick Service Water meeting CAT (NOT MET) System is considered a salt I/Il/Ill. There is no water system and thus would be impact to the 5b 6-15 in the category with the highest application.
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| failure frequencies. Note, using as an example 6' Service Water piping, the segmentbased methodology (Section F.8.1) corresponds to a frequency of 5.8E-06/year, which would correspond to about 2 1/2 feet of piping based on the 'spray' failure frequency (100 gpm or less) of sea water service water
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact pipe per EPRI TR-1 013141.
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| Thus, use of the segment-based approach is questioned since it results in underestimating the pipe failure frequency, particularly for the Service Water system which is an important contributor to Internal Flooding scenarios. It is recommended, as it was by the previous F&O, that the pipe failure frequency methodology be updated.
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| RS 10-03, Section F.5.1 and The plant specific nature of the potential plant systems Resolution of this F.5., discuss the use of plant impacts on flooding frequency was evaluated based on the finding evaluated specific piping configuration and type of water (fresh, salt, etc.), length of pipe (plant specific the plant specific walkdowns to determine flood - room specific), valve rupture and maintenance induced nature of potential initiating event frequency. floods was evaluated in Section F.8 of RSC-1 0-05. plant systems Additionally, generic failure impacts on rates were used for pipe break The BSEP historical flood data is included in section F.1 .1 of flooding IFEV-A6 frequency. RSC-10-05. The EPRI pipe failure rate used BSEP data frequency. This (CAT I) where appropriate to determine the pipe failure rates used finding is It did not appear that plant- (EPRI TR-1 013141 - reference 145 in RCS-1 0-05). sufficiently 3-14 specific information was resolved for the gathered with respect to the SR to be flood LIKELIHOOD. Specifically, assessed as there was no operating meeting CAT IlIll1.
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| experience related to water There is no hammer or material condition of impact to the 5b fluid systems at Brunswick. application.
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| - Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact Documentation of SR IFQU-A2 Table F.B.1 maps flooding initiating events to impacted Resolution of this does not provide a link between rooms. Tables F.5 and F.6 of RSC-10-05 map components finding mapped the flooding IE and the to flood zones (rooms). Table F.14 of RSC-10-05 maps the flood sources, equipment failed for that specific propagation of flood through the flood zones (and rooms). flooding initiating flooding initiator. By inspection events, and flood of the model, equipment and propagation to the IFQU-A2 system fault trees have been appropriate modified to include flood- rooms.
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| (NOT MET) induced failures. No finding isThis 1-2 documentation of the events or sufficiently systems impacted is included in resolved for the the flooding analysis SR to be (maintained as a separate assessed as spreadsheet, see IFQU-B1). meeting CAT Thus, it is not possible to fully I/ll/1l1. There is no verify if the modifications were impact to the 5b done correctly, application.
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| The analyses of all HRA are 1) The flooding HRA's now use a CBDTM method, Resolution of this documented in the HRA which is used for other BNP specific HRA's. finding converted Calculator and were not readily 2) The CBDTM method was used, and alarms were the flooding HRAs available for implemented. System alarms were integrated to the CBDTM review. Based on the time sparsely under the consideration that the leak may method, available for HRA Calculator not be severe enough to reach the setpoint. Sump implemented
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| ]FQU-A5 review, the following issues and alarms were used in all cases.,a alarms, and (NOT MET) comments are 3) Cues are listed in the HRA calculator, "cues" ta. implemented and provided. This should not be 4) Operator discussion was implemented and documented 2-0 considered an all inclusive list documented per BNP-PSA-034 operator given the time restraints on the Time available is presented in tables F.A. 1, F.A.2, F.A.3, discussions. This review, and F.A.4. For events with spray, the analysis does not finding is credit any HRA's for mitigation or recovery for equipment in sufficiently While the HRA analyses did the path of the water spray, resolved for the include many important factors SR to be used to determine a human assessed as
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| _______error meeting CAT
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding [Resolution 5b Impact probabilities (HEPs) associated I/li/Ill. There is no with isolating flood events, the impact to the 5b following issues were judged to application.
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| potentially impact the calculated HEPs:
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| a The methodology used for calculating the HEPs for the flooding termination was the annunciator response methodology. There was no discussion in the flooding documentation regarding the use of, or acceptability of the use of, this methodology. Flooding HRA documentation should be enhanced to include a discussion of the methodology, and the justification thereof, for calculating HEPs.
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| * The use of the annunciator response methodology did not appear to account for all expected annunciators. For example, for large turbine building floods, it does not appear reasonable that only one annunciator would be received.
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| Under the large turbine building flood conditions, it would be o
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding [Resolution 5b Impact expected that the loss of circulating water pumps and the plant transient would cause multiple alarms.
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| Therefore, this HEP could be significantly under-estimated as the annunciator response methodology modeling only assumed that one alarm would be received.
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| * Since the detection and diagnosis of a flooding situation is critical to flood mitigate, a detailed discussion of the detection systems and alarms available for diagnosis is necessary to ensure a realistic HRA. Neither the flooding documentation, nor the HRA calculator files, contains a complete description of the cues available to the operator for flood detection.
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| Therefore, the accuracy of the HEP for flood termination is questionable.
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| * It does not appear that operator interviews or talk-throughs were used to determine the floodincj
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution ] 5b Impact HEPs. Interviews with operators and their response to various alarms would provide invaluable insights into the flood response and would provide for a more realistic approach.
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| *It is not clear that the "time available" value used in the HEP was appropriate for some scenarios. The time available used was the time to which the flood would reach the height of critical equipment.
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| However, the time of the flood to reach the flood isolation valves, in some scenarios, may be time limiting. Additionally, the impact on timing if damage is due to spray, vice submergence, must be considered. The analysis should be more specific regarding which valves need to be isolation, their locations, and the ability to isolate before the flood scenarios cause them to be unisolable.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact No consideration is documented BNP-PSA-035 section F.13 discusses the impact of flooding Resolution of this for the impact of flooding on on the internal events HRA. finding operator actions modeled in the documented the HRA which require operator BNP-PSA-034 and BNP-PSA-035 discuss the flooding impact of flooding actions in areas subject to alarms associated with credited operator actions. Timing, on internal flooding. There is no discussion including discovery, was broken down to the limits events, flooding in the documentation of how the presented in BNP-PSA-035 section F.13. alarms associated flood would be discovered and with credited thus no discussion of the time it BNP-PSA-035 section F.9, discusses the effects of the flood operator actions, IFQU-A6 would take plant personnel to on mitigation equipment. and the effects of (NOT MET) start to respond. There is no the flood on documentation or discussion of mitigation 6-13 this in Section F.11. No equipment. This discussion of whether or not the finding is mitigating equipment listed in sufficiently table F.20 would be affected by resolved for the the flood itself. SR to be assessed as meeting CAT I/Il/Ill. There is no impact to the 5b
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| ____________application.
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| Direct effects were included. Impacts of spray are discussed in section F.10 of RSC Resolution of this The only indirect effect included 05. Equipment considered for spray impact is listed in Table finding discussed was submergence. F.21 of RSC-10-05. Effects of High Energy Line Break the impacts of (pressure, humidity, condensation, temperature and pipe spray events and I FQU-A9 whip) are discussed in section F.1 1 of RSC-10-05. high energy line (NOT MET) breaks. This finding is 1-33 sufficiently resolved for the SR to be assessed as meeting CAT
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact I/Il/Ill. There is no impact to the 5b application.
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| There is no evidence that this The impacts of internal flooding on large early release Resolution of this requirement was addressed. frequency were evaluated during the LERF cutset reviews finding analyzed IFQU-A10 The flooding analysis did not performed as part of Revision 10 of the Quantification the impact of (NOT MET) appear to analyze any impact Calculation (BNP-PSA-030). The review found that internal flooding on the on the LERE model due to the flooding was a contributor to LERF and that the flooding LERF model.
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| 6-14 flooding initiators, hazards were being captured appropriately in the model. There is no impact to the 5b application.
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| The report RSC 10-05 does not Table F.B.1 maps flooding initiating events and source pipes Resolution of this contain the information that lists to impacted rooms. Tables F.5 and F.6 of RSC-10-05 map finding mapped what equipment is damaged for components to flood zones (rooms). Table F.14 of RSC the flood sources, each pipe failure scenario, thus 05 maps propagation of flood through the flood zones (and flooding initiating there is no documentation rooms). events, and flood available for review for this propagation to the IFQU-B2 aspect of defining the plant appropriate (CTIl/l) equipment subject to damage rooms. This (A IIl) from each flooding initiator. This finding is 6-0 information may be contained in sufficiently the database used in performing resolved for the the flooding analysis, but it SR to be could not be confirmed. assessed as meeting CAT 1/11/Ill. There is no impact to the 5b application.
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| IFSN-A2 The flooding analysis (as Flooding analysis in the HRA Calculator implement the Resolution of this (NOT MET) documented in RSC 10-05) alarms used to tell operators that there is a flood occurring finding included discusses sumps and drains, in the system. These alarms are in the control room the use of
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact 1-25 curbs, spray shields, and (described in BNP-PSA-035) and redundant alarms are flooding alarms in watertight doors. No mention is available on the Radwaste panel, there is sufficient alarms the HRA made of flood alarms (or other for the operator. Turbine building floods with operator calculator. This information that would alert actions do not appear in the higher cutsets, and pose finding is operators of the flood), blowout negligible risk. sufficiently panels or HVAC dampers. As resolved for the flood screening is performed SR to be based on assumed operator assessed as intervention, the lack of meeting CAT information concerning alarms I/Il/Ill. There is no impacts the ability to accurately impact to the 5b assess the probability of application.
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| successful flood termination RSC 10-05 does not identify any Alarms are documented in the HRA calculator for the floods Resolution of this automatic flood isolation that have operator actions. Isolation features and timing are finding included features, No operator discussed in BNP-PSA-035. the use of indications are discussed. flooding alarms in However, the qualitative the HRA screening of flood areas calculator. This evaluates successful operator finding is IFSN-A3 action to isolate any flood sufficiently (NOT MET) assuming that the flood is resolved for the immediately detected. The flood SR to be 1-20 documentation needs to indicate assessed as how each flood would be meeting CAT detected and the time to identify I/Il/lll. There is no the occurrence of the flood impact to the 5b needs to be factored into the application.
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| HEP calculation. As the current evaluation is non-conservative,
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| ____________this SR is assessed as not met.____________________________
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # ] _Finding ______[Resolution 5b Impact _
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| The methodology is considered Impacts of spray are discussed in section F.10 of RSC Resolution of this sufficient and-robust for 05. Equipment considered for spray impact is listed in Table finding assessed identifying flood damage due to F.21 of RSC-10-05. the impact of submergence. EPRI TR- spray events and 101 9194 recommends that at Effects of High Energy Line Break (pressure, humidity, high energy line least a 10 foot radius from a condensation, temperature and pipe whip) are discussed in breaks. This pipe be considered for impact of section F.1 1 of RSC-10-05. finding is spray due to pipe failure. Many sufficiently industry PRAs assume up to 30 resolved for the feet for spray impact. While SR to be spray due to water falling down assessed as propagation pathways has been meeting CAT I/I1.
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| assessed for impact (or lack There is no thereof) on SSCs, there is no impact to the 5b IFSN-A6 evidence that this was done for application.
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| (NOT MET) any equipment. Equipment in flood zones such as upper 6-16 levels of the reactor building would likely be subject to this failure mechanism.
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| The R.G. 1.200 qualification for this SR requires that a qualitative assessment of pipe whip, jet impingement, humidity, condensation and temperature concerns in the flooding analysis. The current flooding documentation does not provide this assessment.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact Section F.1.3 of RSC10-05 Section F.4.8 provides basis for not considering drain paths, Resolution of this states that flows through drains wall penetrations, cable trays, and HVAC ducts as they are finding provided a were considered, but there is no insignificant compared to the flood propagation through basis for not discussion concerning drain door, stairwells, and gratings, considering some paths as a possible propagation potential path between rooms in the propagation Auxiliary Building basement nor pathways. This IFSN-A8 the basis for not considering finding is (CAT I) such paths. Discussion of drain sufficiently paths not included, Propagation resolved for the 1-26 through wall penetrations, cable SR to be trays, and HVAC ducts do not assessed as appear to be considered. As meeting CAT I1.
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| only propagation through doors, There is no stairwells, and gratings are impact to the 5b considered, the requirements for application.
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| Cat II are not met.
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| While multi-unit flooding was F.2.1 .2 dismisses the propagation of floods in the turbine Resolution of this considered, all scenarios were building due to "massive" amount of water required to fail finding improved
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| -screened out on the basis of the doors between zones. Section F.1.3 assumes that the flooding physical barriers that are roll up doors in the turbine building are pushed open once documentation to assumed to prevent flood the water has reached 4 inches on the 20ft elevation, address multi-unit ISFN-A1 1 propagation between units. flooding (NOT MET)
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| | |
| ==Reference:==
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| BNP-PSA-035, Rev. 4 scenarios. This IFEV-A4 However, the turbine building finding is (NOT MET) has complete communication sufficiently between unit 1 and unit 2. resolved for the 1-27 Therefore, water can spread SR to be from one unit to other and vice assessed as versa. The flood scenario is meeting CAT described more detail in Section I/lI/Ill. There is no F.2. 1.2 of the notebook. In impact to the 5b addition, turbine building roll-up application.
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact doors on elevation 20' do not appear to have been considered
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| - as a propagation path between units. So, it appears that the screening of all inter-unit turbine building floods may not be correct.
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| Brunswick Flooding PRA credits Drains are not credited as mitigation paths in RSC-1 0-05. Resolution of this operation of drains for mitigating Drains are mentioned as a possible mitigation path for the finding removed flooding events. Per EPRI TR- EDG rooms, but that is only listed as one of the possible credit for drains 1019194, there is wide paths which included doors and stairwells, as flood variability in modeling on this mitigation paths.
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| IFSN-A13 issue, with widespread industry This finding is (NOT MET) practice of not crediting the sufficiently functioning of drains due to the resolved for the 6-18 high probability of sump pump SR to be failures and clogging of drains, assessed as There is no discussion in RSC meeting CAT 10-05 concerning the reliability 1/11/111. There is no of the drains as a mitigating impact to the 5b
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| _______system for flooding, application.
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| Flood sources were screened if Flooding alarm locations are discussed in BNP-PSA-034 Resolution of this more than 8 hours is required to and BNP-PSA-035. finding included reach a one foot depth in an For flooding initiators where operators have time to isolate the use of area. However, there is no the flood, indications are discussed with their respective flooding alarms in IFSN-A14 discussion of whether indication entry in the HRA Calculator. the HRA (NOT MET) is available to identify the flood calculator. This or if isolation can be performed finding is 1-28 given the flooding is occurring, sufficiently resolved for the SR to be assessed as meeting CAT I1.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact There is no impact to the 5b application.
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| Table F.5 of RSC 10-05 lists all Information has been corrected, flooding analysis has been Resolution of this of the potential flooding sources, updated. Fire protection piping described in Table F.3 of finding corrected however one of the major RSC-1 0-05. Pipe break frequencies for Fire Protection Pipe information assumptions is that all fire (by Pipe Identification Number) are included in Table F.16 of concerning the protection sprinkler systems are RSC-1 0-05. fire protection IFSO-A1 dry-type systems.SD-41 system. This (NOT MET) contains a listing of fire finding is protection systems and most sufficiently 1-1 are wet type systems. Since the resolved for the inclusion of fire protection could SR to be significantly alter the screening assessed as and scenario development, this meeting CAT is considered to be not met. 1/11/Il1. There is no impact to the 5b application.
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| Pipe breaks and valve body Gasket and expansion joint caused flooding is now covered Resolution of this failures were the only failure in section F.12 of RSC-10-05. Human induced mechanisms finding added mechanisms identified. Although are covered in section F.8.3 of RSC-1 0-05. gasket and plant experience includes a expansion joint IFS O-A4 gasket failure, failures of caused flooding (N_ E) gaskets, expansion joints, etc. as well as human are not discussed. Therefore,inue SR is not met as all required mechanisms to 1-2 mechanisms are not the model. This considered. finding is sufficiently resolved for the SR to be
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact assessed as meeting CAT 1/11/111. There is no impact to the 5b application.
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| The temperatures and Section F.6 of RSC-10-05 describes the characteristics Resolution of this pressures of condensate, (temperature and pressure) of the various flood sources. finding adequately feedwater, and nuclear service Volumes of flood sources provided in Table F.9 of RSC-1 0- described the piping are not considered. The 05. Flood flow rates provided in Table F.12 of RSC-10-05. characteristics of flow rates for circulating water flood sources.
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| IFSO-A5 piping are based on one pump Circulating water piping based on 4 pump flow - see Table This finding is (NOT MET) instead of all four. This SR is F,12 of RSC-10-05, sufficiently judged to be not met. resolved for the 1-23 SR to be assessed as meeting CAT I/1l/Ill. There is no impact to the 5b application.
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| Most of the level of detail Table F.12 maps flood sources to room and Table F.B.1 Resolution of this required to relate the flood maps flooding initiating events to impacted rooms. Tables finding mapped sources to actual rooms and F.5 and F.6 of RSC-1 0-05 map components to flood zones the flood sources, piping and components is (rooms). Table F.14 of RSC-10-05 maps propagation of flooding initiating located in a vendor database flood through the flood zones (and rooms). events, and flood (NSOTMET that is not part of the propagation to the (FNOTAET documentation file. Therefore appropriate (NOT MET) this SR is not met. rooms. This finding is 1-24 sufficiently resolved for the SRs to be assessed as meeting CAT
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| _________________________________I/Il/Ill. There is no
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| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b impact impact to the 5b application.
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| Documentation of the internal IFSN-B1 - Discussion of flooding scenarios enhanced in Resolution of this flooding work was not conducive section F.4 of RSC-1 0-05. finding improved to supporting PRA applications, documentation maintenance and upgrade, or IFSN-B2 (c) - Assumptions for submergence, spray, and the PRA peer review. Much of the temperature and other flood effects given in sections F.1 .3, model analysis.
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| information was contained in a F.9, F.1 0 and F.1 1 of RSC-1 0-05. This finding is IFSN-B1 contractor flood database that sufficiently (NOTI MET) was not captured in the IFSN-B2 (d) - General screening criteria for analysis in resolved for the IFSN-B2 Brunswick document system sections F.1.2 and F.1.3 of RSC-10-05 and specifics are SRs to be (NOT MET) and which was not accessible given in greater detail in sections F.2 and F.3 of RSC-10-05 assessed as IFSO-B1 for review, on an area by area basis. meeting CAT (NOT MET) There is no I/Il/Ill.
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| IFEV-B1 As noted in IFSN-B1, the flood IFSN-B2 (e) - Flooding scenarios considered are identified impact to the 5b (NOT MET) scenario development in section F.2 of RSC-1 0-05. Screening of flooding application.
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| IFEV-B2 documentation is not believed to scenarios is performed in section F.3 of RSC-10-05.
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| (NOT MET) provide all of the information Retained flooding scenarios are described in section F.4 of IFQU-B1 needed to fully describe the RSC-1 0-05.
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| (NOT MET) scenario development process.
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| IFQU-B2 In addition to other items noted IFSN-B2 (f) - Pipe break, spray and human induced (NOT MET) in the IFSN-A SRs, items initiators were added to the model, Tables F.38 and F.39 of (c),(d),(e), and (f) of SR IFSN- RSC-10-05 identify components that were modeled with 1-31 B2 need to be included as part flooding failures. The flooding initiators are listed in annex of the documentation. Also, a F.C of RSC-1 0-05.
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| listing of the specific components assumed to be Tables F.5 and F.6 of RSC-1 0-05 identify flooding failed in each flood area needs components by flooding zone and Tables F.38 and F.39 of to be provided. Discussion of RSC-1 0-05 identify PSA components that were modeled propagation pathways could with flooding failures. Flooding pathways are described in
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # ]Finding IResolution 5b Impact have been at a greater level of section F.4 of RSC-1 0-05 for important flood areas. Flood detail. propagation on room by room basis is provided in Table F.14 of RSC-10-05. Assumption on behavior of doors is Documentation of risk-important provided in section F.1.3 of RSC-10-05.
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| components within each flood area appeared incomplete in the IFQU-A5 - Section F.13 of RSC-10-05 gives screening HRA walkdown notes. Documentation for flooding events to determine importance of flooding of characteristics of flood events. Tables F.26 and F.27 identify operator actions that pathways between zones was would be impacted by flooding events. Additional HRA weak and/or incomplete, often development for flooding scenarios is included in BNP-PSA-indicating only the first pice of 0034.
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| equipment to fail as flood waters rise. The level of detail in the IFQU-A1 - System failures due to piping failures given in walkdown notes was considered section F.9 and Tables F.19 and F.20 of RSC-10-05.
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| weak. Documentation and discussion of flood pathways other than doors was limited.
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| Assumptions on door behavior, such as pressures doors would withstand or flood propagation rates through door gaps, were not documented.
| |
| Operator actions to be modeled in the Human Reliability Analyses for Flooing were not clearly identified and described, including documentation of alarms that would be considered for specific flooding scenarios.
| |
| Review of plant operating experience pertinent to these HRA analyses were not documented. It was not possible
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact to verify proper development of the HRA Calculator Files, as discussed under IFQU-A5. This contributed to being unable to verify that IFQU-A5 requirements were met.
| |
| As discussed in IFQU-A1, the Flooding Analysis does not provide information concerning modeling of system/component failures due to pipe failures in that system, as opposed to failure due to flooding and flood propagation.
| |
| Documentation of the general philosophy and approach used for the flooding analysis can be improved.
| |
| Parameter values were selected 1) LOSP curves were updated to more modern composite Resolution of this with regards to the PRA curves. Appendix C tables 4.7-2, 7.7-5. finding updated Standard's requirements for HR 2) Component failures have been updated in the LOSP curves and and DA. Consideration of documentation. See BNP-PSA-049 Appendix C component severe accident conditions upon failures. There is LE-E1 these parameters is provided in no impact to the (CAT I/Il/Ill) Appendix M, or in some 5b application.
| |
| instances Appendix C, of the 1-19 BNP-PSA-049 notebook.
| |
| Section G of LE notebook captures the human error modeling, and incorporated the general methodology approach
| |
| ___________used in Level 1 HRA.
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact However, the data values documented in BNP-PSA-049 were developed during a previous PRA update. It appears that some values may need to be updated to be consistent with changes in the Level 1 data. For example, aSP recovery values (such as ACP1XHE-MN-OFFE) are not consistent with the current OSP recovery curve (and LOSP is now categorized by type of OSP failure as opposed to a composite value). On the other hand, changes in component failure data appear to have been updated in the Level 2 trees.
| |
| However, the documentation does not indicate that the values shown in BNP-PSA-049 have been superseded.
| |
| There is very limited discussion Uncertainties are discussed in detail in BNP-PSA-075. Resolution of this of the impact of variability / Table 1, items 228 to 271. This outlines both the finding discussed sensitivity in time to core uncertainties associated with the LERF basis and the the uncertainties LE-G5 damage amongst different limitations for use, associated with (NOT MET) methodologies upon potential the LERF basis applications in Appendix C of and the limitation BNP-PSA-049. While limitations for use. This 6-2 of the quantification process are finding is discussed in BNP-PSA-030 sufficiently Sec.3.6, that discussion is not resolved for the
| |
| _______pertinent to SR LE-G-5. It is SR to be
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact concluded there is not sufficient assessed as discussion of the limitations of meeting CAT the LERF analysis that could I/Il/Ill. There is no impact different applications, impact to the 5b thus this documentation application.
| |
| __________requirement is NOT MET.__________________________ ________
| |
| Dependency analysis was Combinations of operator actions among the top 95% of Resolution of this performed on the identified HFE cutsets were analyzed, and the larger bulk of dependencies finding analyzed combinations (see BNP-PSA- addressed to ensure a correct value. See BNP-PSA-034, combinations of 034 and associated section E.4.1 .5 operator actions spreadsheets). The dependency and assessment approach used dependencies to appears to be appropriate. In ensure correct developing recovery rules to be values. There is applied to the cutsets, maximum no impact to the combinations of 3 HFEs were 5b application.
| |
| included. Any cutsets with greater than three HFEs that QU-C2 meet the recovery rule criteria are recovered to a minimum joint HFE of 1 E-6 (and often 3-9 higher). As a result, there are cutsets that contain more than three HFEs that are being recovered to a higher frequency than may be warranted (either because one or more of the additional HFEs may be independent of the others, or because the joint HFE probability is still above the floor value of 1 E-6 (and often higher).
| |
| As a result, there are cutsets that contain more than three
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact HFEs that are being recovered to a higher frequency than may be warranted (either because one or more of the additional HFEs may be independent of the others, or because the joint HFE probability is still above the floor value of 1E-6 and hence could be reduced further). This conservatism appears to increase the calculated CDF/LERE by at least a modest amount.
| |
| It is stated in BNP-PSA-030 that The initial cutset review is documented in Section 3.7 of Resolution of this the top 200 cutsets have been BNP-PSA-030. The cutset reviews are completed by a finding reviewed a reviewed. However, to panel of PRA personnel. The full range of cutsets was sample of cutsets determine if there are logic reviewed strategically to understand the most significant over the full range problems buried deeper down in cutsets, ensure that the results are reasonable and of significance the cutsets, A sample should be complete. The panel performs the review by analyzing the and improved QU-D1 taken of significant cutsets or first 200 cutsets in detail, analyzing approximately 100 documentation.
| |
| (NOT MET) sequences. The correct cutsets from the 201 through non-significant at random This finding is definition of significance is intervals, and analyzing a small number of the non- sufficiently 2-7 stated in the quantification significant cutsets for completeness. Accident Sequence resolved for the notebook. The review should cutsets were also reviewed to ensure cutsets are correct SR to be include a sampling of cutsets or and consistent with system model and success criteria and assessed as sequences over the full range of to ensure significant and non-significant cutsets are showing meeting CAT significance (i.e., top 95% of up appropriately. Section 3.3 and 3.4 provide a detailed I/Il/Ill. There is no cutsets contributing to review and description of significant cutest (i.e. individually impact to the 5b DF/LERF).
| |
| C______ contributing more than 1% to CDFI/LERF). application.
| |
| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact Section 3.9 of BNS-PSA-030 Revised the BSEP Quantification calculation BNP-PSA-030 Resolution of this compares BNS PSA results to include an enhanced similar plant review. Section 3.9 finding revised the against other units. However, contains two tables, one that compares very broadly quantification there is no analysis of Brunswick 1 &2 against other BWRs and another that calculation to contributors as required for Cat compares Brunswick 1&2 to Hatch and Browns Ferry in include the I1.In addition, no references are more detail. The broad review looks at analysis of provided for the other PRAs that differences/similarities in plant design, ODE and LERF. The contributors and are compared to (e.g., are these in depth review examines the individual initiators to COF. references for the QU-D4I IPE results, or current results, Most of the Brunswick model initiators are within an order of review of similar (A ) etc.) magnitude and are considered to compare similarly. The plant PRAs. This 4-7 largest outlier in the comparison is the contribution of internal finding is flooding initiators to the Brunswick core damage frequency. sufficiently The large difference is due to a Brunswick specific design resolved for the characteristic were a break in a condensate pipe traveling SR to be through the cable spread rooms causes an internal flood that assessed as fills the room before mitigation activities can be successful. meeting CAT IlIll1.
| |
| There is no impact to the 5b application.
| |
| Failures of the F032A and 1-B21-F032A/B are stop check valves that are upstream of Remains F032B outboard feedwater the F010OA/B. The likely hood of two check valves failing in unmodeled due to check valves are not modeled, series is improbable. During HPCIIRCIC operations in the low probability.
| |
| The system notebooks state plant, these valves are shut by MOV thus sealing them into HPCI (F032A) or that such failure is considered the closed position unlike normal check valves which may RCIC (F032B)
| |
| SY-A13 "improbable". However, the stick in the open position due to fouling of the hinge system (CAI/l/ll) model does include the failure of mechanism. The PRA models them with a fails to remain unreliability is two inboard check valve F010OA to open BE due to the credit given for feed in HPI.oresf 4-5 reopen. magnitude higher than the check valve failure probability. The unmodeled failure mode is bounded by the uncertainty
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| | |
| Table 2. BSEP Internal Events and Internal Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding jResolution 5b Impact of the system unreliability.
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| There is no impact to the 5b application.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # [Finding JResolution f 5b Impact Generic MSO scenario 2e The identified MSO scenarios were re-evaluated using Resolution of this appears to be inadequately deterministic and thermal hydraulic methods for individual finding re-evaluated dispositioned. The scenario MSOs and combinations of MSOs. The results of the re- the MSO scenarios identified in NEI 00-01 is a drain evaluation concluded that the individual MSOs and mentioned. These of the vessel, while the rough combinations of the MSOs did not result in a failure of MSOs remain calculations evaluate this as credited components, addition of new initiating events or screened from essentially (word in Attachment 3 a change in accident sequences. These MSOs remain as inclusion in the EPRA of the component selection screened from inclusion in the FPRA model. The analysis model. There is no report) a depletion of the of these MSOs has been updated in Attachment 3 of the impact to the 5b ES-B2 suppression pool. The loss is component selection calculation, BNP-PSA-085. application.
| |
| (CAT 1,1II) estimated as a 200 gpm loss, ES-D1 (CAT 1/l11/1ll) which can be an issue either a) long term for inventory, or b) in PRM-B9 combination with other small (CAT I/Il/Ill) losses (See NEI 00-0 1 for PRM-C1 guidance on combining MSOs).
| |
| (CAT I/WI/ll)
| |
| Generic Scenario 2d appears to also be a possible long-term 1-2 issue (with multiple seal failures),
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| or an issue in combination with other small losses.
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| Scenario B21-2c (Main steam drain line) includes an evaluation of flow size listed as 0.03 square inches based on a single flow path. However, multiple drain line openinqs are possible.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The MSIVs spurious operation Given the MSIVs are normally open during power Resolution of this appears to be modeled as a operation, MSIV spurious opening or failure to close finding added two new ES-A5 failure of containment isolation cannot be a fire-induced initiating event. MSIV LOCA accident (CAT I, II) under gate 1 Si. This spurious sequences. There is ES-A6 operation does not appear to be However, two new MSIV LOCA accident sequences were no impact to the 5b (CAT I, II) modeled as either an initiating created to model a fire-induced post-trip MSIV spurious application.
| |
| ES-B2 event or LOCA, or showing to opening or failure to close (MSO-B21-2b). These (CAT I, II) impact RCIC/HPCI operation, sequences do not credit HPCI or RCIC and include the ES-D1 loss of the condenser.
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| (CAT I/Il/lll) Most BWR FPRAs include MSIV failure to close or spurious re- This has been documented in Section 3.3.1.4 and opening as a large or medium Attachment 3 of the component selection calculation 1-6 LOCA, given downstream BNP-PSA-085.
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| opening of TBVs or other large steam line valves.
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| The FPRA modeling does not A detailed review of the fire induced initiating events was Resolution of this
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| -include mapping of multiple point performed, with particular attention to those initiating finding performed a estimate initiating events to events identified by the Peer Review, and was detailed review of fire ESA1 specific equipment. This includes documented in Section 3.3.1.4 and Attachment 8 of the induced imitating E-l the following lEs: Loss of offsite component selection calculation BNP-PSA-085. The events. Logic for fire (NOT MET) power, Inadvertent opening of review found that all initiating events had been adequately induced LOOP was ES-A2 SRV (%1"1-S), Loss of DC Power addressed except for fire induced LOOP. Based on the added to fault tree (CAT I/Il/Ill) (%1TDC1A1, %1T_DC1B2), sensitivity study cited in F&O 1-8, no change was made where appropriate.
| |
| ES-A3 Loss of Switchboard for the loss of DC power, because DC initiators are This finding is (NOT MET) (%1T_DC1A, %1TDC1 B), and adjacent to response model events in the fault tree. Logic sufficiently resolved FQ-A2 loss of AC Bus (%1TEEl, for fire induced LOOP was added to the fault tree where for both SR ES-Al (CAT I/1l/Ill) %1TE_E2). In essence, these are appropriate. Inadvertent SRV opening was removed from and SR ES-A3 to be treated as a plant transient (in this IAN"*G005 which is present under IAN^G178 as assessed as meeting case, an MSIV closure event) documented in Rev 2 of the change log, Attachment 9 of CAT I/ll/Ill. There is 1-8 followed by a subsequent failure BNP-PSA-085. Attachment 3 of the component selection no impact to the 5b of the equipment. calculation BNP-PSA-085 documents MSOs that were application.
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| evaluated as possible initiators but determined not to be A sensitivity case was requested creditable.
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| for the loss of DC Power Al and
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # {Finding jResolution j 5b Impact ioss of offsite power to determine the possible impact on the COD P.
| |
| The results show some differences in the cutsets and the CCDP results, mainly due to actuation logic (applied under the IE logic), restart logic, and failure of CRD. Overall, the CCDP following the IEs is slightly higher than assuming the subsequent failure of the equipment.
| |
| Significant tracing was performed of the logic for each IE. In most cases, the IE logic was ORed with the equipment logic.
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| However, there were exceptions.
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| For example, for the inadvertent SRV opening; gate IAN1 G178 (HEADER A ISOLATED AND NOT RECOVERED) included the IE but not the equipment logic for SRV opening. Another example:
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| Gate #U13 (S2 LOCA OR SORV WITH ONE OR MORE SRVS FAILING TO RECLOSE) includes SRV logic above, but only for 2 or more SRVs. As a result, the single SRV opening for the IE is not included under this logic.
| |
| Review of LOOP logic indicated several locations where consequential LOOP was not
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact included; although the logic included in most cases other fire logic such as MSIV closure, or other assumed fire lEs.
| |
| Feedwater and HPCI overfeed is The applicable impact of Feedwater and HPCI overfeed, Resolution of this not included in the FPRA as initiator events, is already appropriately modeled. finding added modeling for possible Fire- documentation to Induced Initiating Events. Because it does not degrade the ability of the plant to justify the current mitigate the resulting transient, Feedwater overfeed modeling approach.
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| (MSO-N21-2ai) was included in the FPRA as an initiating This finding is event which is subsumed within the Turbine Trip initiator, sufficiently resolved ES-Al This is consistent with the treatment of initiating events in for SR ES-Al to be (NOT MET) the Internal Events model (BNP-PSA-032) and is assessed as meeting ES-A4 supported by the results of the MSO Expert Panel review. CAT 1/11/111. There is (CAT I/Il) Generically, NEI-00-01 does not list MSO-N21-2ai as no impact to the 5b FQ-A2applicable to BWR4s, noting that steam-driven feedwater application.
| |
| FQA2 pumps may not be a concern, and (upon review) the MSO (CAT I/Il/Ill) Expert Panel concurred.
| |
| Likewise consistent with the treatment of initiating events 1-9 in the Internal Events model (BNP-PSA-032), the MSO Expert Panel did not considered a plant trip to be a creditable result of a spurious HPCI operation (MSO-E41-2u). However, the possible effect of spurious HPCI operation (MSO-E41 -2u) on the ability of the plant to mitigate an otherwise initiated transient was considered.
| |
| In particular, during a postulated spurious HPCI operation (MSO-E41-2u), the high RPV water level signal may not
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact isolate the steam inlet valve, but Operating Experience suggests that the turbine would over speed on low quality steam and mechanically trip at some point prior to the RPV water level actually reaching the steam lines.
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| Consequently, RPV water level is not anticipated to induce a concurrent RCIC failure. However, since the available Operating Experience does not specify the RPV water level at which the steam quality is assured to cause a turbine trip, the RPV water level is identified as a source of uncertainty.
| |
| Documentation to justify this position has been added to Section 3.1, Attachment 3, and Attachment 8 of the
| |
| ________________________component selection calculation, BNP-PSA-085.
| |
| SY-B5 Instrumentation included in the The power supplies for the instrumentation credited for Resolution of this (CAT I/il/Ill) FPRA that affects HFES are operator actions, as identified in BNP-PSA-084, have finding added the SY-A6 listed in Calculation BNP-PSA- been added to the FPRA. The revision is documented in power supplies for the (NOT MET) 084, Revision 1, attachment 4. the component selection calculation (BNP-PSA-085) instrumentation S-9 This attachment provides a model change log. Power supplies were already included credited for operator (CAT I/l I/Ill) comprehensive list of instruments in the component selection to support other modeled actions. This finding affecting each of the modeled equipment. is sufficiently resolved ES-Cl HEPs in the PRA. for SR SY-A6 to be (CAT I/Il/Ill) assessed as meeting PRM-B9 However, the power supplies for CAT I/Il/Ill. There is (CAT I/Il/Ill) the instrumentation added to the no impact to the 5b FPRA model is not included in the application.
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| FPRA logic 1-10
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding ]Resolution 5b Impact Change Package BNP-0122 The BSEP trip assessment in Change Package BNP- Resolution of this includes a list of plant areas, and 0122 (Attachment 10 of BNP-PSA-080) was updated finding updated the an evaluation of a possible plant using additional insight of targets in each fire BSEP trip trip for each area. The categories compartment/zone. Targets identified for both the safe assessments using include near certainty plant trip shut down and the fire probabilistic assessment were insights of targets in (1.0), reduced likelihood trip (0.1) considered. The likelihood of a plant trip due to fire was each fire and plant trip not likely (0.01). evaluated on an area basis and considered both the compartment. There In discussions with the engineer equipment located in the area and the equipment that is no impact to the 5b who developed this list, the could be affected by traced cables that traverse the application.
| |
| assessment was based on area. The evaluation considered the possible effects on judgment, which including some the plant given a fire in the area but not the likelihood of a consideration for the likelihood of fire in that area or the assumed status of the system in the fire, some consideration of the the FPRA. To accommodate operator discretion, no area possible damage of a fire, and the was assigned a conditional trip probability of zero, even if equipment in the area. However, the area contained no equipment important to plant PRM-B4 the judgment did not include a operation.
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| (CAT I/Il/Ill) review of cables and equipment impacted in each area.
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| The results show that areas 1-14 impacting safety busses (which would result in a likely rapid plant shutdown), are estimated to shut down the plant 10% of the time, while impacts from spurious operation (e.g., SRV openings, MSIV closures, etc.) are not accounted for. Additionally, the base assumption of all fires causing a plant trip, loss of feedwater, loss of condenser vacuum, and MSIV closure (e.g.,
| |
| no cable tracing for these initiating events) is not applied.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding JResolution [ 5b Impact The transient ZOI approach was 1)The turbine buikldng was re-examined for a transient fire ln the NFPA 805-based on the 75th fire versus the representing a 317 kW 98% HRR, as documented in Safety Evaluation, the 98% fire. As a result, the transient Attachment 16 of Revision 1 of BNP-PSA-086. To NRC staff found the scenarios were impacted as support the use of a lower HRR in specific areas and use of the lower HRR follows: consistent with the clarified guidance in Section G.5 of to be acceptable NUREG/CR-6850, as endorsed by the NRC in a letter because the a
| |
| : 1) Scenarios were rnot identified in dated June 21, 2012 (i.e., ADAMS Accession Number characterization of the areas where the cable trays were ML12171A583), an evaluation was performed as past plant-specific above 6 feet, but below the zone documented in BNP-PSA-086, Attachment 25. That transient combustible of influence for a 317 kw fire evaluation included a review of plant records for the violations was (height depends on location). performance of OFPP-01 3, Transient Fire Load provided indicating Evaluation, covering a period of two years. From those that the lower HRR is
| |
| : 2) Area for the ZOI was limited. records, a sample of 10 were selected for more in-depth achievable and For example, in the cable room, review, which consisted of plant walk downs, operator because an improved the area for each transient interviews, interviews with the responsible engineer, and procedure was FSS-A1 scenario was typically 3' x 3', a verification of plant wide training. In the Service Water established to provide (NOT MET) versus a longer area which may Building, where plant records repeatedly indicated confidence that all fire impact a particular cable tray. elevated combustible material, the walkdown included the loads will be limited to Again for this area, several cable inspection of the actual burnable material to determine the lower HRR.
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| 1-19 tray runs are 30' or longer, where the lower HRR to be reasonably realistic and bounding. This finding is the area assumed for a larger ZOI Where the higher HRR was determined to be more sufficiently resolved would be something like 30' x 7.' reasonably realistic and bounding for a particular fire for SR FSS-A1 to be compartment, the evaluation also considered whether the assessed as meeting
| |
| : 3) Areas, such as the Battery fire compartment contained equipment or targets that CAT I/Il/Ill. There is Rooms have no identified might be impacted by the higher HRR and whether further no impact to the 5b transient scenarios. evaluation would be required. Where necessary, the application.
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| evaluation identified specific fire compartments to be subjected to future administrative controls. Subsequent to the evaluation described in Attachment 25 of BNP-PSA-086, the Fire Protection Program System Health Reports covering the last three years were reviewed for violations of the transient combustible controls. These reports document and evaluate both internal plant records
| |
| ('e..g., Condition Reports) and NRC inspection records
| |
| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions (Resolution SR/F&O # jjFinding 5b Impact (e.g., by the NRC resident inspector or during triennial inspections). Condition Reports concerning events that occurred outside the Global Plant Analysis Boundary (GPAB) or during an outage were screened as not applicable to the Fire PRA. Condition Reports concerning events that occurred in the Turbine Building, where the higher HRR was used, or in Fire Compartments that were qualitatively screened in the Fire PRA, were also screened as not applicable to the use of a lower transient HRR. The evaluation also eliminated from further consideration those events where the corrective action was likely sufficient to preclude recurrence and those events that would likely not be considered a violation under Fl R-NGGC-0009. Because FlIR-NGGC-0009 includes both provisions for supporting the reduced HRRs credited in the Fire PRA and a list of specific combustible materials that have been pre-evaluated to be acceptable within those constraints, none of the violations so identified could have resulted in a transient fire exceeding the reduced HRRs credited in the Fire PRA. For the remaining Condition Reports, the evaluation considers what administrative controls are applicable under FIR-NGGC-0009 and qualitatively what effect on risk a fire might have if those controls should fail. The use (post-transition to NFPA 805) of the more proactive FIR-NGGC-0009, is credited with limiting the placement of transient combustibles and ignition sources near equipment and cables unless a specific evaluation is performed using a 317 kW 98% HRR.
| |
| : 2) The floor area applied for each transient scenario is based on the identified target set. The minimum applied transient foot print is 3'x3'.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact
| |
| : 3) All plant areas were assigned transient ignition source(s). If the transient ignition source did not damage any significant targets, no risk increase would be recorded from that potential fire source. This was the case for the Battery Room.
| |
| Transient scenarios are identified Transient and other fire impacts on sensitive electronics Resolution of this using a ZOl assuming cable were incorporated into the EPRA in Revision 5 of BNP- finding justified the damage only. No damage to PSA-080, as documented in Attachment 54. Otherwise, current modeling equipment appears to be consistent with the guidance in H.2 of NUREG/CR-6850, approach. This assumed for any area. all of the ZOIs are based on cable damage. It would be finding is sufficiently FSS-A1 very conservative to assume equipment damage based resolved for SR ESS-(NOT MET) For example, a transient fire in on the same ZOI. In most cases the equipment is Al to be assessed as the battery room was not shielded to some degree by a steel enclosure, such that meeting CAT 1/1l/Ill.
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| developed where the transient internal damage would be minimal from an external There is no impact to 1-20 damages or ignites the batteries, source. Also, exclusion zones exist to limit placement of the 5b application.
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| which is near the floor. Another unattended transient ignition sources next to MC~s /
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| example is there are no scenarios energized equipment (ref. 0FPP-014 and FIR-NGGC-located between 1CB and 1 CA, 0009).
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| where damage to both cabinets may occur.
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| HR-El No new actions were identified in Following the Peer Review a detailed HEP was Resolution of this (NOT MET) response for the Fire. developed to provide a more realistic evaluation of a finding developed an HR-E2 Discussions with BSEP remote shutdown following control room abandonment. A HEP for the control (NOT MET) operators, the ASSD procedures review of ASSD-01 and ASSD-02 identified key operator room for the will be used for shutdown given a recovery actions and related system interfaces. A proper implementation of HR-E3 fire and damage to ASSD understanding of system operation within the context of a ASSD-01 and ASSD-(CAT I,11II/ll) equipment. For example, ASSD- fire scenario was obtained during focused talk-throughs 02. This finding is HR-E4 01 will call for shutdown outside and operator interviews. The results of the HRA including sufficiently resolved (CAT I,11II/ll) of the control room in ASSD-02. the operator interviews, are documented in Attachment 10 for SR HR-El, SR
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact FSS-B1 For other areas, there are specific of BNP-PSA-084, Revision 2. HR-E2, and SR HRA-(CAT I/Il/Ill) ASSD procedures. ASSD-05 was Possible conservatisms associated with not modeling A2 each to be HRA-A2 reviewed for fire in Unit 1 RX'Bld other ASSD actions are not considered to be significant. assessed at meeting (NOT MET) North. This procedure includes CAT 1/11/111 and for SR specific recovery actions and HRA-D1 to be HRA-C1 manual actions, including for assessed as meeting (CAT I, II) example operation of the SRVs CAT II. There is no HRA-D1 from the RSP. Neither the control impact to the 5b (NOT MET) room evacuation actions or the application.
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| local manual actions were identified or reviewed as a part of 1 -24 the fire PRA. As a result, the FPRA results are conservative.
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| For example, the top cutset for the cable room could be recovered using a control room evacuation action.
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| The control room abandonment A detailed Human Event Probability (HEP) has been Resolution of this HEP for habitability scenarios developed for the control room abandonment scenario finding developed an uses a CCDP of 0.1 and CLERP due to habitability concerns, as documented in HEP for the control HR-G1 of 0.01, without detailed analysis Attachment 10 of BNP-PSA-084. An evaluation of the room abandonment (CAT I) or support. These values may be various key operator actions contained in the scenario. This finding FS-2 conservative or non-conservative, abandonment procedures was performed using the is sufficiently resolved (CTI) depending on the scenario CBDTM/THERP (CT, I) (including equipment damage) Calculator. Themethodology contained evaluation uses in the HRA Safe Shutdown timing for SR HR-G1 assessed to be as meeting HRA-C1 and timing. studies and feasibility analysis. The specific training and CAT I1. There is no (CAT 1,II) frequency of training was evaluated as well as a detailed impact to the 5b No detailed timing, feasibility, review of the procedure. Significant equipment failures application.
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| review of training, review of were also considered in the determination of the CCDP.
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| 1-26 procedures, or detailed task The HEPs resulting from the HRA calculator evaluations analysis was documented in the were then placed into an event tree with supporting top FPRA. logic to determine an overall HEP.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact MOO fire scenarios do not include When an open MOO fire was modeled, it is conservatively Resolution of this propagation from one MOO stack assumed that the entire MOO is failed and all targets finding justified the to another. NUREG/OR-6850 within the ZOI of the MOO are treated as failed by the fire. assumption that includes a propagation model, When the cabinet remains closed, the entire MOO is breaching with where propagation is assumed failed and the fire is assumed to remain confined to a subsequent normal FSS-A1 following a 10-15 minute delay single stack. Except that a certain probability of arcing - fire growth is possible (NOT MET) (depending on the opening). induced breaching is postulated for closed MO~s with even for closed subsequent normal fire growth. Reference FAQ 14-0009. MO~s. This finding is The BWROG methods (not sufficiently resolved 1-30 approved) includes a probability for SR FSS-A1 to be of propagate and an approach for assessed as meeting limiting the number of cabinets OAT 1/11/111. There is considered in propagation, and no impact to the 5b an approach for determining the application.
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| HRR.
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| * Location factors w(e.g., wall New walkdowns were performed for transient sources in Resolution of this effects) were not included in HRR the Turbine Buildings. The new walkdowns increased the finding accounted for calculations for transients. HRR and accounted for wall effects. These have been wall effects for incorporated into the Fire Scenario Data calculation transients and
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| * Review of the documentation (BNP-PSA-086). Use of wall effects in other areas has incorporated 211lkW shows that ceiling jet treatment been added as an uncertainty in the calculation. HRR for pumps into FSS-C1 was not performed. the FPRA. This The identification of targets from is based on a zone (OAT I)
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| * The 75th and 98th percentile influence (ZOI) determined the source HRRof using finding resolvedis for sufficiently SR ESS-HRR assigned for pumps accepted and approved methods. Where secondary fire 01 to be assessed as 1-2 (electrical fire) are from Case # 7 growth is expected, The ZOl treatment is conservatively meeting OAT II.
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| 1-2 for motors, BIN 14 (69 kW) in lieu extended to the ceiling. The treatment of ceiling jets There is no impact to of from Case # 6 for pumps, BIN would only be addressed when more detailed fire the 5b application.
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| 21 (69 and 211 kW). modeling is applied. This does not typically apply to transient sources since the overall transient analysis is based on virtual sources and does not contain the specific inputs that would be needed to justify the applicability of detailed analysis.
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| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution 5b Impact New target wakdowns were performed using 211 kW for the HRR for pumps and were incorporated into Revision 5 of BNP-PSA-080, as Attachment 44.
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| A screening value for rated The BSEP fire quantification calculation has been revised Resolution of this barrier probability of 1 E-2 was and the screening of HGL Multi Compartment Analysis finding performed the FSS-G2 applied. This may not be has been performed in accordance with NUREG/CR6850. screening of HGL (CT11/1) bounding depending on the The screening value of 0.1 was used on the exposing MCA in accordance (CTI/lIl) features of the barrier (doors, compartment to screen out compartments from the MCA with NUREG/CR6850.
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| penetrations, dampers). analysis. The results of the revised Multi-Compartment There is no impact to 1-34 Analysis is documented in the quantification calculation the 5b application.
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| BNP-PSA-080.
| |
| QU-B2 Truncation in the CDF and LERF The truncation approach has been changed in Rev 1 of Since the process for (NOT MET) was varied, based upon the the quantification calculation (BNP-PSA-080) in response establishing truncation QU-F2 CCDP/CLERP. For example, to this F&O. Scenarios are now run at an effective limits does not (CAT 1/l1/Ill) CCDP of 1 .0 uses a truncation of truncation of 1 E-09/yr for CDF and 1E-1 0/yr for LERF demonstrate that the U-B 1.0, while a CCDP of 1E-03 uses which is more than four orders of magnitude below the overall model results QU-B3ino E-7 vrltereutn D n LR ln oas cnegS UB a(rnaioNf1OT7 process using the ones vral (OME) h rslig D nMLRElnttTl.)ovreS UB run will continue to be FQ-B1 results in difficultly running assessed as NOT (CAT I/1l/I1l) FRANC at a very low cutoff. MET. However, the FQ-F1 very low effective (NOT MET) A review of the truncation levels truncation (i.e.,
| |
| was performed. Hundreds of the relative to the sequences have truncation within resulting CDF and 1-36 a factor of 100 or less of the LERF plant totals)
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact CCDP. Several of these provides reasonable sequences were re-run, and the assurance that no new CDFs were compared to the significant accident original CDFs. Changes in the sequence was results vary from about 5% to as inadvertently much as 25%. Many of the eliminated. This sequences affected are in the top finding is sufficiently 25 fire sequences. resolved for both SR QU-B2 and SR EQ-Fl Additionally, a large number of to assessed as scenarios are listed with zero meeting CAT I/Il/Ill.
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| COD P. When these were re-run There is no impact to with lower truncation values, the 5b application.
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| cutsets were generated. This can be important for scenarios with higher ignition frequencies.
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| LE-G2 A quantitative evaluation of LERE A quantitative evaluation of parametric uncertainty for Resolution of this (NOT MET) uncertainty was not included in both CDF and LERF was performed as documented in finding provided the LE-F3 the final results. The uncertainty EVAL EC 296040, including a State of Knowledge required quantitative (NOT MET) quantification was performed for Correlation covering fire ignition frequencies, non- evaluation of UN-A CDE results only. suppression probabilities, conditional failure probabilities, uncertainty. This (NOT MET) Assumptions and key areas of and fire bins. finding is sufficiently uncertainty did not include resolved for SR LE-EQ-E1 discussion of LERF, other than G2, SR LE-F3, SR (NOT MET) the use of a simplified LERF UNO-Al, SR EQ-El, EQ-Fl value for control room and SR FQ-F1 to be (NOT MET) abandonment. assessed as meeting CAT 1/11/111. There is no impact to the 5b 1-38 ____________________________________application.
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| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The BSEP EPRA roadmap In the process at BSEP, Fire Protection/NSCA develops Resolution of this CS-Al indicates that the methodology to and maintains the cable selection and circuit analysis finding justified the (CAT I/ll/Ill) identify additional cables uses data. These data are then referenced as inputs to the current methods. This CS-A3 same process for PRA circuit Component Selection and Quantification FPRA finding is sufficiently (CAT 1/11/111) analysis as for the deterministic calculations. This process and associated results are resolved for SR CS-CS-Cl Safe Shutdown circuit analysis. easily reviewable, has been peer reviewed multiple times C1 to be assessed as (NOT MET) Reference FIR-NGGC-01 01. for our other sites and found to be acceptable. There is meeting CAT 1/11/111.
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| However, there is no separate no requirement to have a separate PRA notebook. There is no impact to notebook for Fire PRA Cable the 5b application 2 Selection to discuss the processes, inputs and results.
| |
| The QLS screening criteria may Revision 2 of BNP-PSA-083 removed FC261 Resolution of this not have been applied (DUCTBANK) from qualitative screening and retained it finding removed appropriately. BNP-PSA-083 Rev, for quantitative analysis. FC261 (DUCKBANK) 1 Section 3.3 documented the from qualitative QL-1 screening criteria as in Raceway target information, cable loadings and floor screening and (CSAT/lll) NUREGICR-6850. Alternate areas for the manholes in the DUCTBANK were collected, retained it for (CTII/l) screening criteria was used to MOS factors were assigned, and transient ignition quantitative analysis.
| |
| QLS-.A2 screen several analysis units. frequencies were determined. There is no impact to (CAT I/Il/Ill) These criteria were based on the the 5b application.
| |
| QLS-A3 judged low risk significance of the With the exception of FC261 (DUCTBAN K), no physical (CAT I/Il/Ill) unit in question. analysis unit was qualitatively screened based on the use QLS-A4 of alternate screening criteria. However, FC295 and (CAT I/Il/Ill) Moreover, FC261 (DUCTBANK) FC345 (i.e., DrywelliTorus, for Unit 1 and Unit 2, QLS-B3 was screened out based on no respectively) were not retained for quantitative analysis (CAT I/Il/lll) equipment while the QLS because no ignition frequency was assigned to the screening criteria need to rule out Drywell/Torus based on the Technical Specifications both equipment and cables. requirements for an inert atmosphere during power 2-3DUCTBANK will contain a large operations. This treatment is consistent with both the Fire number of cables and low risk Hazard Analysis in the (U)FSAR and the Safe Shutdown contribution is not expected. Analysis in which no fire is postulated in or analyzed for the Drywell.
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| BSEP team responded as follows:
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # jFinding Resolution j 5b Impact
| |
| 'FC261 (DUCTBANK) is not a typical fire compartment (i.e., '. ..a well-defined enclosed room...').
| |
| As described in Attachment 3, FC261 (DUCTBANK) is a network of underground conduit in pre-cast concrete cable trenches.
| |
| Rather subsuming FC261 (DUCTBANK) into FC263 (with certain 'yard' locations), FC261 (DUCTBANK) was separately identified during plant partitioning to promote clarity in communication with legacy plant fire protection programs. Because of its design, no transient fire was postulated for FC261 (DUCTBANK). As described in Attachment 3, no equipment is located in the FC261 (DUCTBANK). And as stated in Section 3.4.2, all cables at BSEP are considered qualified self extinguishing and non-propagating.
| |
| With no creditable ignition source, there is no fire risk. Therefore, it was considered appropriate to qualitatively screen FC261 (DUCTBANK) consistent with the stated intent of the general task objective described in Section 4.3.1 of NUREG/CR-6850.'
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact However, it is not expected to have absolute zero fire ignition frequencies in the ductbanks since these enclosed areas could be open for maintenance during outages and transients could be left there unnoticed, even the transient materials from plant startup. On the other hand, 100%
| |
| qualified self extinguishing and non-propagating cables may not be realistic. Cables used for lighting and other not modeled system functions may exist in the ductbanks, which may not be
| |
| _____________ qualified.
| |
| SY-C1i The system notebooks (i.e., A discussion of fire impacts was added as Section 3.7 of Resolution of this (NOT MET) calculation BNP-PSA-062) have BNP-PSA-062, Revision 11, and included fire-related finding added a SY-A2 been updated for MOR1 1 from changes (e.g., disposition of MSOs) to the FPRA and discussion of fire (NOT MET) which the FPRA was sources of model uncertainty and related assumptions. impacts and fire S-2 subsequently developed. The related Changes to the SY2 system notebooks (i.e., system notebooks.
| |
| (NOT MET) calculation BNP-PSA-062) will be This finding is SY-A3 further updated to incorporate sufficiently resolved for (NOT MET) fire-specific changes to the SR SY-C1, SR SY-A2, SY-A4 model. SR SY-C2, SR SY-A3, (CAT I) and SR SY-A6, to be SY-A6 However, the system analysis assessed as meeting (NOT MET) supporting requirements included CAT 1/11111l and for SR PR-9 in SY-A2, A3, A4, A6, C1 and C2 SY-A4 to be assessed (CARMll/ll have been determined to be not as meeting CAT IlI/ll.
| |
| (CT11/1) met with the current There is no impact to PRM-C1 documentation, which was
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution j 5b Impact (CAT 1/11/111) typically performed by updating the 5b application.
| |
| the system notebooks to reflect all fire-related changes.
| |
| 2-8 An example of information not included from SY-A2 includes:
| |
| COLLECT pertinent information to ensure that the systems analysis appropriately reflects the as-built and as-operated systems.
| |
| Examples of such information include system P&IDs, one-line diagrams, instrumentation and control drawings, spatial layout drawings, system operating procedures, abnormal operating procedures, emergency procedures, success criteria calculations, the final or updated SAR, technical specifications, training information, system descriptions and related design documents, actual system operating experience, and interviews with system engineers and operators.
| |
| See other referenced SRs for other information not included in the FPRA documentation.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The system notebooks A discussion of fire impacts was added as Section 3.7 of Resolution of this (calculation BNP-PSA-062) have BNP-PSA-062, Revision 11, and included fire-related finding added a been updated for MOR1 1 from changes (e.g., disposition of MSOs) to the FPRA and discussion of fire SY-C3 which the FPRA was sources of model uncertainty and related assumptions. impacts and fire (NOT MET) subsequently developed. The related changes to the PRM-C1 system notebooks (calculation system notebooks.
| |
| (CAT I/Il/Ill) BNP-PSA-062) will be further This finding is updated to incorporate fire- sufficiently resolved for specific changes to the model. SY-C3 to be assessed 2-9 However, the sources of model as meeting CAT I/Il/Ill.
| |
| uncertainty and related There is no impact to assumptions are not documented. the 5b application.
| |
| PRA items that were assumed The requested sensitivity is on items considered always The current failed for the component selection failed in the Fire PRA. This treatment represents a assumption is QU-E4 are listed in BNP-PSA-085 Rev. 1 conservatism in the calculated Fire CDF. Of these, the identified as a (CAT 1/11/111) Section 4 and Table 4. largest effect is likely the assumption of loss of feedwater conservatism in the UNC-A1 This treatment is similar to for each scenario. calculated Fire CDF.
| |
| (NOT MET) treatment of unknown locations The finding is PRM-B10 for equipment that do not have sufficiently resolved (CAT I/li/Ill) cable-routing completed, for SR UNC-A1 to be Sensitivity studies should be assessed as meeting performed to investigate the risk CAT I/Il/Ill. There is 2-10 importance of these failed no impact to the 5b systems/functions. application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact BNP-PSA-080 Section 4.5.3, Section 3.2.3.3 of BNP-PSA-083 documents Resolution of this Non-Suppression Probability, consideration of the applicability of using generic non- finding justified the documents the methods used for suppression data based on an outlier review of plant fire current methods. This calculation of non-suppression bridge experience. To support the FPRA Peer Review a finding is sufficiently probabilities. Generic NSP and review of the then-current one year period (i.e., 2011) of resolved for SR FSS-unavailability are applied from System Health Report information was performed and D7 to be assessed as NUREG/CR-6850. No outlier indicated no "outlier behavior" for the Fire Detection and meeting CAT I1.
| |
| review is performed, and no plant Suppression Systems. This focus on a 1-year period There is no impact to specific data are used to update provided an overview of the most current system the 5b application.
| |
| the unavailabilities. performance, measured against specific parameter/attributes, and helped to confirm the effectiveness of preventative and corrective maintenance.
| |
| FSS-D7 During the NFPA 805 RAI response, the most recent (CAT I) three years of System Health Report information were also reviewed and found to show sustained acceptable performance levels, again with no "outlier behavior" noted 2-14 for the Fire Detection and Suppression Systems.
| |
| Currently, system performance is monitored and maintained at a high level as part of the System Health Reporting and System Notebook processes. Outlier behavior with respect to system availability would be evident to the system engineer and plant management through the health data (available for the previous 12 months), which indicates overall Excellent (Green) performance. Post-transition, the assessment of system performance is part of the NEPA 805 Monitoring Program, as described in procedure FIR-NGGC-0130.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SRIF&O # [Finding Resolution J 5b Impact The note of SR FSS-D8-1 states: Accommodation of area specific features and scenarios is Resolution of this Fire detection or suppression assured for fire suppression and detection system finding reinforces the system effectiveness depends on, through correct application of fire protection design correct application of at a minimum, the following: 1) standards such as NFPA 13- Standard for the Installation fire protection design system design complies with of Sprinkler Systems, and NFPA 72- National Fire Alarm standards which applicable codes and standards, Code. In each case careful selection of occupancy ensures that physical and current fire protection classification and hazard identification is applied. This features and the fire engineering practice, 2) the time ensures that physical features and the fire sources sources contained in available to suppress the fire prior contained in a given area are properly protected to a given area are to target damage, 3) specific achieve the desired performance results. Ceiling properly protected to features of physical analysis unit configurations, blockage of agent application by design achieve the desire and fire scenario under analysis features and adequate coverage for the hazards present performance results.
| |
| (e.g., pocketing effects, are a direct function of code compliance. Code There is no impact to blockages that might impact compliance is further assured by detailed evaluation in the 5b application.
| |
| plume behaviors or the "visibility" the NFPA 805 Transition report Table B-i, th rough the FSS-D8 of the fire to detection and use of or reference to Code Compliance calculations such (I/Il/Ill) suppression systems, and as 0FP-1038, Rev. 1, Code Compliance Evaluation NFPA suppression system coverage), 13 (Reactor Building), 1976 and 1983 Ed. or 0FP-1 031, and 4) suitability of the installed Rev. 0, Code Compliance Evaluation NFPA 72E, 1984 2-15 system given the nature of the fire Ed. for BSEP.
| |
| source being analyzed.
| |
| In light of BSEP fire scenarios, above item 1 should be considered met although not evident in documentation. Timing (item 2) is considered in detailed NSP calculations were carried out in the spreadsheet files BN P_EVAL_Ul1_CDF.xls, BNP_EVAL_UI_LERF.xls, BNP_EVAL_U2_CDF.xls, and BNP_EVALU2_LERF.xls.
| |
| .1____________________________________________________________ .1________________________
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact However, above items 3 & 4 were not addressed.
| |
| BNP-PSA-086 R0 Section 10.0 Section 4.13 of Calculation BNP-PSA-095 evaluated the Resolution of this states that fires resulting in treatment of smoke damage, as discussed in Appendix T finding evaluated the significant smoke production of NUREG/CR-6850, and the possible effects on the potential for smoke could cause additional damage quantification results in the BSEP FPRA. Appendix T of damage to EPRA beyond the heat based zone of NUREG/CR-6850 limits the equipment vulnerable to short equipment on a influence target sets collected. term smoke damage to medium and high voltage qualitative basis and FSS-D9 However, targets that are switching or transmission equipment, and lower voltage incorporated the (CAT I) susceptible to smoke damage instrumentation and control devices, results of the have not been identified and are The BSEP FPRA currently accounts for smoke damage assessment into the currently not evaluated in this consistent with the guidance in Appendix T of definition of fire calculation. Therefore, this SR is NUREG/CR-6850 by failing the entire electrical bus or scenario target sets.
| |
| 2-6 considered not met for CC-lII. panel where the fire is postulated. This accounts for any This finding is smoke damage generated inside the panel. sufficiently resolved for SR FSS-D9 to be assessed as meeting CAT li/I/ll. There is no impact to the 5b application.
| |
| A review of FRANC model files The FPRA database query qSourceBE 2a(source) Resolution of this showed that some HGL scenarios which identifies the failed components for the individual finding modified the FQ-A1 (whole room burnout) have less scenarios was modified to use the same FSSPMD existing mapping (i/Il/Ill) affected components than some mapping table (TRoutingFireZone) that is used for tables for use in both individual scenarios in the same generating the HGL component failures (Reference BNP- individual scenarios fire compartment (FC) modeling a PSA-080 Rev 1). and HGL scenarios.
| |
| 2-19 single ignition source and targets There is no impact to in its ZOI. the 5b application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding [Resolution 5b Impact For example, in Unit 1 CDF FRANC model, FC212 scenario BHGL has 64 affected components while scenarios FC212_4612 B75 and B98 have 112 affected components.
| |
| On the other hand, some other scenarios have significantly more affected components in HGL scenarios than individual scenarios in the same Fo.
| |
| Discussion with BSEP PRA team indicated that different mapping tables have been used for HGL scenarios and individual ignition source scenarios. Conservatism may exist in the generation of mapping tables for individual scenarios. However, non-conservatism could exist for the HGL scenarios if the different mapping tables do not cover all the cables / equipment that are affected by the fire-induced failures.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact FC238_5010 and similar fire Metal water spray shields are provided over several fire Resolution of this scenarios are expected to have initiators in the Diesel Generator Basement. Specifically finding verified plant significant SBO contributions. A these metal shields are installed over the EDG Excitation design, addressed review of mapping table and Voltage Source PT & Reactor and EDG Excitation potential failures, and excluded events and altered Current XFMR Phase A, B, & C as documented in enhanced the event tables did not show the drawing F-1319. The spray shields are BSEP plant documentation failure of DG Breaker spurious configuration and are maintained via controlled drawings. concerning the metal failures excluded, which is also The shields are designed to prevent water spray water spray shields.
| |
| evident in FRANC affect impingement onto the transformers described above and There is no impact to components. However, it is noted per controlled drawing examination and plant walkdown the Sb application.
| |
| that the IGN is set to 0. they also provide a non-combustible barrier to the development and passage of a damaging fire plume BSEP PRA team responded that: above these transformers. Based on walkdowns the
| |
| 'During review of cutsets following construction of these shields is sufficient to prevent direct preliminary quantification, several passage of a damaging fire plume to targets located EQ-Al scenarios were identified as directly above the protected transformers.
| |
| .(I/il/Ill) significant contributors to plant risk. Review of these scenarios The primary concern with a fire in the subject identified significant transformers (i.e., sources 5010, 5011, 5012, 5013 5014, 2-20 conservatisms in the initial data 5015, 5016 and 5017) is development of a fire plume that inputs that were causing would impact cable trays routed above the spray shield.
| |
| unrealistic risk results. As part of The design of the spray shield is such that the plume this review, it was identified that would be forced to follow a circuitous path prior to the fire size for sources 5010 impingement on the target cable tray. The worst case fire through 5017 were initially expected to develop in the fire initiators would be a 69kW characterized as 211 kW fires fire based on the 98% HRR for dry-type transformers, when a more detailed Ref. NUREG/CR-6850.
| |
| examination of the equipment showed that the sources should All of the cables located in the Diesel Generator be characterized as 69 kW fires. Basement are IEEE 383 qualified; therefore their damage In addition, a shield above the temperature is 625°F and damaging heat flux is 11 kW/m 2 .
| |
| sources was identified. The target cable trays are located above the EDG Consideration of either of these transformer spray shields therefore damaging two facts would result in temperatures must be exceeded at the spillage points of
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # jFinding ]Resolution J 5b Impact consequences for the fires that the spray shield to be deemed capable of damaging the would be much less severe than cable trays or a damaging radiant heat flux radiated from the initial walkdown information the spray shield. Based on review of the spray shield indicated. This information is design and plant walkdown of the initiator/target documented in change package configurations it is judged that the spray shields installed BNP-0182 and BNP-0176 (See above these transformers will prevent thermal damage to the BNP-0176 change package the target cable trays, thus damage resulting from a directory of BNP-PSA-080 calc transformer fire need not be postulated.
| |
| for pictures of these sources).
| |
| Because the quantification Continued maintenance of these spray shields is ensured process was nearly complete, by plant documentation and credit for these spray shields explicitly incorporating the as a radiant/plume shield for raceways located above the information from the change EDG transformers is documented in the fire PRA packages into all input calculation.
| |
| calculations would have resulted in a significant administrative burden to revise the calculations.
| |
| Therefore, to simulate the correct effects within the quantification calculation, the scenarios were assumed to be equivalent to the first target tray having a solid bottom as per BNP- 0176 and the scenario event frequency was set to zero for scenarios FC238_5010 B75 and FC238_5010 B9B. It is assumed that cable trays with solid bottoms will prevent damage to cables for ignition sources with HRR 69 KW or less based on the discussion provided in section Q.2.2 of NUREG/CR-6850.'
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding _Resolution j 5b Impact A test fire model run with FDS was also constructed to demonstrate the adequacy of above engineering judgment. As a result, the technical basis supporting the treatment of the identified scenarios is considered acceptable. However, the following issues should be addressed:
| |
| : 1. The documentation in BNP-0176 should be enhanced to include engineering judgment as discussed above instead of a simple assumption that metal cover above the cabinet is sufficient in preventing fire damage to targets above the cover.
| |
| : 2. The BSEP team stated that the metal cover is part of the design basis. This fact should be verified and documented in fire PRA.
| |
| : 3. The potential failure of the metal cover should be addressed.
| |
| May need to credit the surveillance / inspection /
| |
| maintenance program to ensure the integrity of this metal cover.
| |
| : 4. Perform sensitivity study or include the failure probability of this metal cover to generate risk insights associated with the assumption associated with this
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions S R/F&O # Finding Resolution 5b Impact metal cover.
| |
| : 5. Revise the fire PRA model to not set ignition frequency to 0 but remove the impacted targets.
| |
| BNP-PSA-080 Section 4.3.4, Fire The listed non-instrument spurious cable failures were Resolution of this Induced Spurious Event analyzed, and probabilities were included in the Fire PRA. finding analyzed the Probabilities, document the Conditional failure probabilities were assigned to the most listed non-instrument methods used for conditional risk significant contributors, causing them to become less spurious cable failures failure probabilities for fire- risk significant and allowing these less risk significant and the probabilities induced circuit failures, contributors to appear relatively more risk significant. were included in the More could have been done, but the iterative process Fire PRA. This Circuit Analysis was performed in stopped when satisfactory results were obtained, finding is sufficiently CF-Al change package BNP-0137 to resolved for SR CF-(CAT I) determine the probability of a In many of the identified cases, failures are in Al to be assessed as CF-B1 spurious operation for various instrumentation, and probability analysis methods are not meeting CAT IlI/ll.
| |
| (CTIl/l) cables, available, and no testing has been done to determine the There is no impact to (CT /I/Il)failure probabilities. Division of failure mode based on the 5b application.
| |
| Risk significant contributors were conditional probability analysis would only serve to add not identified (quantification was additional uncertainty to the failures.
| |
| 2-22 complete later in the process) and utilized thus cannot met the The current analysis is conservative in that for cases capability category CC-Il. where specific conditional probabilities have not been developed, failure or spurious operation is give a For example, the Unit 1 CDF probability of 1.0.
| |
| importance results include the following spurious events for which conditional probabilities have not been developed:
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding jResolution f 5b Impact HPC1 PPS-SA-N12A_TPRESSURE SWITCH E41 -N012A SPURIOUSLY ACTUATES HPC1 PPS-SA-N12CTPRESSURE SWITCH E41-N012C SPURIOUSLY ACTUATES RCI 1TME-HI-N021B_TTEMPERATURE ELEMENT E51 -TE-N021 B SPURIOUS OPERATION RCI 1TME-HI-N022B_TTEMPERATURE ELEMENT E51-TE-N022B SPURIOUS OPERATION RCI 1PPS-SA-N012ATPRESSURE SWITCH E51-N012A SPURIOUS OPERATION RCI 1 PPS-SA-N012C_TPRESSURE SWITCH E51-N012C SPURIOUS OPERATION HPC1 PPS-SA-N 12B_TPRESSURE SWITCH E41-N012B SPURIOUSLY ACTUATES HPC1 PPS-SA-N12D_TPRESSURE SWITCH E41 -NO12D SPURIOUSLY ACTUATES SRVlISRV-CO-F01 3G TNON-
| |
| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions SRIF&O # ]Finding -
| |
| Resolution 5b impact ADS SAFETY RELIEF VALVE B21-FO13G SPURIOUSLY OPENS RHR1 MDP-SA-COO2CTRHR PUMP E11-COO2C SPURIOUS START DUE TO FIRE RCI1 PPS-SA-NO12BTPRESSURE SWITCH E51-NO12B SPURIOUS OPERATION RCI 1PPS-SA-NO12DTPRESSURE SWITCH E51-NO12D SPURIOUS OPERATION HPC1 PPS-SA-N17ATPRESSURE SWITCH E41-NO17A SPURIOUS OPERATION HPC1 PPS-SA-N17BTPRESSURE SWITCH E41-NO17B SPURIOUS OPERATION SwS1 PPS-SAP129LTPRESSURE SWITCH PS129 SPURIOUS OPERATION FAILS LOW ISOLATES HEADER Note that if the instrument spurious operations above are not caused by a hot short, detailed circuit analysis is likely not needed. However, the valve and pump spurious operation
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact would likely benefit from additional analysis.
| |
| BNP-PSA-080, Attachment 17, The Seismic-Fire Interaction Analysis report has been Resolution of this Section 5 documents the failure updated to address this F&O. Sections 5.3.3, 6.4.3, and 9 finding updated the or spurious operation of detection have been updated to address updates since the IPEEE. Seismic-Fire and suppression systems. Interaction Analysis Flooding, habitability and life A list of suppression systems that were modified from dry- report. There is no safety concerns are also pipe to wet-pipe systems was determined from DBD-62. impact to the 5b addressed, but only through Using ESR 94-00345, it was confirmed that the previous application.
| |
| reference to the IPEEE. However, flooding analysis conducted for the plant remained valid no update to this evaluation is for these suppression systems. Therefore, the SF-A2 provided. During the walkdown, it modification of these systems did not introduce any new (CAT I/Il/Ill) was noted that some changes in flooding concerns, and the conclusions from the IPEEE SF-A3 the fire suppression system had evaluation remain valid.
| |
| (CAT I/Il/Ill) recently occurred, including changing some systems from dry The Seismic-Fire Interaction Analysis report has been to wet-pipe systems. updated to address this F&O. Sections 5.2, 5.4, 8, and 9 3-4 have been updated to address common fire pump suction The following have also not been piping.
| |
| specifically addressed:
| |
| Discuss seismic vulnerability of The piping between the diesel and motor driven fire any common fire pump suction pumps is not seismically qualified. Based on drawing piping. A common suction for review and relevant site documents, a single break in the both the electric and diesel fire suction piping from the Fire Protection Water Tank or the pump is provided from the DWT would not result in the loss of both fire pumps due to 300,000 gallon storage tank. the presence of isolation valves. If multiple breaks were to
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution Sb Impact Failure of this line can result in occur due to a seismic event, water supply to both fire failure of both fire pumps. pumps could be compromised. DWT suction piping is not considered vulnerable as it is routed underground in some areas.
| |
| BNP-PSA-080, Attachment 17, The Seismic-Fire Interaction Analysis report has been Resolution of this Section 5.2 discusses common updated to address this F&O. Sections 5.2 and 9 have finding updated the cause suppression failures of the been updated. Seismic-Fire fire water system. The common Interaction Analysis cause failure of gaseous The Unit 1 and 2 HPCI fire compartments each contain an report. There is no suppression system (002 and automatic 002 suppression system. Each system is impact to the Sb Halon) is not discussed. supplied by two banks of C02 supply tanks, designated application.
| |
| the main and reserve banks. These supply tanks are No discussion is provided in located outside the Reactor Building that they serve. Unit SF-A3 regards to establishing redundant 1 HPCI Fire Compartment FC-RB1-2 is served by the SR3 supply of fire water or gaseous main and reserve banks in Fire Compartment HCB1, and (CAT I/Il/Ill) agent supply. Unit 2 HPCI Fire Compartment FC-RB2-2 is served by the main and reserve banks in Fire Compartment HCB2.
| |
| 36Plant procedures should Each set of main and reserve banks serves only the 3-6 specifically address availability of automatic suppression system for the adjacent Reactor redundant fire water and gaseous Building.
| |
| agent supply in case of loss of the main supply of fire water or Based on the close proximity of the main and reserve normal gaseous agent supply. banks for each system, and their location in a non-seismically qualified fire compartment, a seismic event could damage both the main and reserve supply banks and cause the 002 system they supply to become inoperative. However, because the supply for FC-RB1-2
| |
| ____________and FC-RB2-2 are separated by a large, open distance,
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding ]Resolution 5b Impact there is no common cause failure that could result in the loss of supply for both automatic 002 suppression systems.
| |
| The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-O1). The Halon suppressant supply for the system in this fire compartment is local, and so common cause failure is not a concern.
| |
| Discussion of the availability and use of alternate water supply was increased in the report. These alternate supply sources include the DWT and Intake Canal, while the alternate pressure source if both fire pumps are unavailable is an external pump truck. If the fire pumps are unavailable, water supply and pressure can be maintained in the fire suppression ring by external pumper truck through yard hydrants.
| |
| Each carbon dioxide system for the Unit 1 and 2 HPCI fire compartments contain a main and reserve supply bank, but no other redundant supply was found for these systems. The Unit 1 and 2 systems do not share a common supply and cannot be cross-tied.
| |
| The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so redundant supply due to common cause failure was not examined.
| |
| Plant procedure 0OP-41 includes procedures used to
| |
| _____________align the fire protection system to alternate water supplies
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&0 # Finding Resolution 5b Impact and an alternate pressure supply.
| |
| There is a selector switch for each 002 system to select between main and reserve banks, but no procedure was found for the use of this selector switch. The operation of this selector switch should be included in a procedure to allow for transfer from the main to reserve bank (or vice-versa) in the event the selected supply bank becomes unavailable. No physical cross-tie or procedure to align one unit's HPCI C02 system to the 002 supply of the other unit was found in this analysis.
| |
| The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so procedures to align a redundant supply due to common cause failure were not examined.
| |
| Justification for partitioning Additional justification/clarification was added to BNP- Resolution of this elements that either lack a fire PSA-083 for the partitioning elements that lack a fire finding added resistance rating or have been rating, especially with regard to the presence of justification for PP-B1 omitted need to be provided for intervening combustibles for open partitioning elements, partitioning elements (CAT I/ll/Ill) the following fire compartments that lack a fire rating.
| |
| PP-B2 (examples only): This finding is (NTMT oFC207 - The east wall has an sufficiently resolved open doorway to FC206 which is for SR PP-B2 to be PP-C3 not justified assessed as meeting (CAT I/Il/Ill) oFC210O/FC211- -fire rated seals CAT IlIll1. There is no that cannot be maintained as fire impact to the 5b barriers application.
| |
| 3-8 *FC238 (DG-1) - This compartment also interfaces with FC244, 245. No justification of partitioning.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding JResolution [ 5b Impact
| |
| * Generic - Block walls are rated for 2 hours per 3.2.2.2, however, the walls column identifies them as 3 hours in most cases. Some cases no rating is provided.
| |
| * FC252 - No justification for unrated block wall - south.
| |
| *FC269, 270, 271, 272 - No justification for open grating and stairwell. The only discussion is that openings are beneficial in preventing HGL. If partitioning is not an issue, then it could be combined as one area.
| |
| Transients or fixed combustible ignition sources and intervening combustibles close to the opening may result in damaging plume temperatures beyond the compartment and/or affect OMAs and fire response.
| |
| * FC274, 275 - compartment above separated by concrete ceiling and open chase. No justifications for open pipe chase, except that it aids in preventing HGL.
| |
| * FC278, 279, 284, 285 - Open stairwell, electrical chase and pipe chase are not justified.
| |
| * FC270 is spatially separated from FC269 by the mezzanine space above HPCI room. No justification has been provided, I _________________________________________________________ I _______________________
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution j 5b Impact i.e. distance, intervening combustibles, combustible free zones, etc.
| |
| Severity factor calculations are Severity factors were applied to every scenario, based on Resolution of this based on generic data per ignition the approved calculation (NED-M/MECH-1 006). finding justified the source and the distance to the However, most had a Severity Factor of 1.0 because the current process. This nearest target (BNP-PSA-080). closest target was within the ZOI of the lowest HRR Bin. finding is sufficiently FSS-C1 Review of the BPNFPRA These distances are based on well documented walk- resolved for both SR (CAT I) database (and associated down results. Sources were typically evaluated for at FSS-C1 and SR FSS-FSS-C4 BNP_EVAL spreadsheet) shows least two HRRs based on the 75% and the 98% 04 to be assessed as (OAT I) that the distance from the ignition percentile fires. meeting CAT I1.
| |
| source to the nearest target is 0 There is no impact to inches for 3779 of the 4907 the 5b application.
| |
| 3-12 sources (including transients).
| |
| Other target distances are mostly few inches from the source.
| |
| Resulting SF is 1.0 for almost all scenarios.
| |
| Unreviewed Analysis Method This analysis method was piloted at HNP and is generally Resolution of this described in FAQ 14-0009, which was being developed finding justified the 3S-Al concurrent with NRC review of the BNP NFPA 805 LAR. current process and The BSEP FPRA calculates SS-A using: In Section 3.4.2.2 of the associated Safety Evaluation, the promulgated the jt Met) 1) A severity factor 0.1, where NRC found the 0.1 cabinet breaching factor to be method to the rest of 90% of the fires are contained acceptable for well-sealed MO~s at BNP because it is the industry through 4-1 within the MOO consistent with available operating experience and is the FAQ process.
| |
| : 2) HRR severity factors are systematically applied to a representative physical This finding is treated independently, similar to configuration. sufficiently resolved
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact other cabinets, for SR ESS-A1 to be In the final version, FAQ 14-0009 conservatively uses a assessed as meeting breaching factor of 0.23. This relatively minor difference CAT 1/11/111. There is does not invalidate the acceptability of the existing no impact to the 5b approach but is considered new data which would be application.
| |
| evaluated for incorporation into the FPRA as part of the normal model maintenance process.
| |
| BSEP cable routing information is All of the cables routed for FSSPMD contain the terminal Resolution of this contained in the BSEP FSSPMD information. In fact, the cable naming includes the finding confirmed that database. This data base termination information. There are some instances where terminal information is contains cable routing information these data have not been repeated in the FROM/TO included for all cables CS-Al10 for the selected cables and fields, however this field not required for BSEP. These in FSSPMD. There is (CAT Il) includes routing information for fields exist because in other plants the termination no impact to the 5b CS-C2 the analysis unit and raceway information must be entered specifically. application.
| |
| (CAT I/llI/Ill) information for the subject cables.
| |
| The database includes treatment of cable terminal end locations for 4-.5 most cables contained in the database. However, several cables were found with no terminal data included.
| |
| No new thermal hydraulic New engineering calculations, thermal hydraulic analysis, Resolution of this analysis was used in the and simulator runs were performed to confirm the finding performed SC-B1 construction of the fire PRA, success criteria previously established by engineering additional analysis to (NOT MET) however, there are several judgment. There was no change to success criteria confirm the success instances where engineering previously modeled that were based on engineering criteria previously PRM-B7 judgment was used to justify no judgment. The component selection calculation (BNP- established by (CAT I/llI/Ill) changes are required in the PSA-085) has been updated to reference the specific engineering judgment.
| |
| existing success criteria, calculations used in determining the success criteria. This finding is sufficiently resolved 4-8 There are several instances for SR SC-B 1 to be where thermal hydraulic analysis assessed as meeting could have been used to replace CAT II. There is no
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact engineering judgment in the impact to the 5b justification or no justification was application.
| |
| found for use of existing success criteria in the internal events criteria: 1) no evaluation of the affects on the thermal hydraulic calculation and or timing was found for MSO C1 1-2e (RPV coolant drain through the SDVvent and drain) for loss of 138,000 gal of suppression pool inventory on accident progression. 2)T23-4U (Spurious opening of torus vent and purge valves) no thermal hydraulic evaluation of long term affects of short term containment failure on long term containment over pressure. C71-lA (ATWS) -
| |
| Justification states that hot shorts may last for up to 11.3 minutes, this may have a significant impact on the thermal hydraulic analysis, this needs to be considered if this timing is used in the justification for exclusion of the MSO.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding IResolution 5b impact In general analytical methods F&O 4-9 seems to confuse expert judgment with Resolution of this were not used in the limited conservative decision making, further confounding the finding justifies the changes for success criteria for issue by labeling the later as engineering judgment and use of expert the fire PRA. All of the analysis suggesting that CATiI prohibits its use. Expert judgment judgment as opposed reviewed has some type of is defined in the standard as "information provided by a to analytical methods.
| |
| engineering judgment included in technical expert, in the expert's area of expertise, based This finding is the justification. on opinion, or on an interpretation based on reasoning sufficiently resolved MSO P41-5e is an example for a that includes evaluations of theories, models, or for SR SC-B2 to be change in the success criteria of experiments." This differs markedly with the example assessed as meeting a credited system which includes cited, MSO P41-5e, in that the MSO involves a limited CAT IlIll1. There is no engineering judgments and or number of possible outcomes (i.e., either the flow impact to the 5b assumptions for the justification. diversion fails the NSW pump or it does not). Assuming application.
| |
| that the NSW pump fails is certainly the more Case # P41 -5e
| |
| | |
| == Description:==
| |
| conservative decision. Citing hard data (e.g., pump SC-B2 Spurious operation (open) of both design capacity or operating flow rate) for the expected (CAT I)
| |
| RHR service water isolation performance of specific equipment as the basis for PRM-B7 (crosstie) valves in a loop may making a conservative decision should certainly not (CAT I/Il/Ill) result in diversion of service water cause the resultant stated assumption to be treated with flow from the RHR heat the same level of scrutiny as the, presumably much exchangers. softer, information based on an opinion formed from the 4-9 evaluation of a theory, model, or experiment.
| |
| PRA Disposition: 'Each nuclear service water pump has an 8,000 gpm design capacity. Each RHR SW heat exchanger has a design flowrate of 8,000 gpm. The RBCCW system is adjusted for a 7,200 gpm flow rate. The RBCCW system only automatically isolates on a LOCA or LOOP signal. Since LOCA's are not considered in a fire PRA, it is assumed that one nuclear service water pump is needed
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SRIF&O # Finding Resolution 5b Impact and aiigned to RBCCW at the time of the fire. It the second NSW pump automatically starts (including discharge valve opening) on low NSW header pressure, a spurious opening of one RHR HX path will be mitigated. If two or more RHR HX paths spuriously open, it will be assumed that both NSW pumps will fail due to run-out. Otherwise, if the standby NSW pump does not start of its discharge valve does not open, only one RHR HX path needs to be spuriously opened to fail the operating NSW pump. In this case the standby NSW pump will also be failed (due to the assumed valve or pump failure).
| |
| The following combinations model this MSO (and SO R43-5j).'
| |
| The BSEP approach of fire The Fire Scenario Data calculation BNP-PSA-086 has Resolution of this FSS-A4 scenario development was to been updated to address this F&O. Section 9.5.2 has finding updated the (NOT MET) evaluate all identified fire sources been updated to include fire propagation. Fire Scenario Data ESS-D1 1 individually. These fire scenarios calculation to include (CAT I/Il/Ill) included the specific cable tray, The database was updated by adding several queries that fire propagation. This FSS-G1 component, and conduit targets create tables which determine the secondary initiator finding is sufficiently (CAT I/I I/lll) for each credible source. within the most limiting ZOI. All other targets that are resolved for SR ESS-However, review of the located above the secondary initiator (larger DIST_V A4 to be assessed as information determined that the value) are then included to be in the same ZOI as the meeting CAT 1/11I/Ill.
| |
| 4-1 1 identified targets included were limiting secondary initiator. This is done by setting the There is no impact to only those within the zone of fields [69 kW], [143 kW], [211 kW], [317 kW], and [702 the 5b application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact influences of the initial source. kW] for all targets vertically above the limiting secondary No additional targets were initiator to match the same fields of the limiting secondary included that were in the zone of initiator in the table [Z Source-Target].
| |
| influence for fire growth scenarios intervening combustibles such as cable trays were in the original zone of influence.
| |
| The assessment used to quantify While conservative, CAT II can be MET if the fire risk is Resolution of this the fire risk for the unscreen bounded. finding justified the analysis compartment used fairly current approach.
| |
| conservative approaches, such Some type of fire modeling was performed for all ignition This finding is as consideration of only the 75% sources and the development of the FPRA included the sufficiently resolved and 98% fires, inclusion of selection and application of either computational or for SR FSS-D3 to be suppression based on damage to noncomputational fire modeling tools consistent with the assessed as meeting the first target and/or manual guidance in Section 11.5.1.7.1 of NUREG/CR-6850. CAT I1. There is no suppression only for time to While there may be conservatisms associated with the impact to the Sb FSS-D3 damage of first target of 15 selection and application of particular fire modeling tools, application.
| |
| (CAT I) minutes. the fire modeling tools described in NUREG/CR-6850 are considered sufficiently accurate, despite any associated No specific fire modeling, conservatisms, for the fire modeling of all physical 4-13 calculations, or analysis were analysis units and scenarios to be sufficiently realistic done of the significant fire units rather than bounding.
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| analyzed in the quantification tasks. More analysis was The selection and application of fire modeling tools was included for the MCR with respect part of a larger iterative process for evaluating high risk to abandonment, however, there contributors. This process involved a team of still significant conservatism knowledgeable individuals with diverse expertise, remaining in the calculations such including fire modeling, circuit analysis, PRA, and plant as the below noted in BPN-PSA- operations. As described in Attachment 39 of BNP-PSA-080 'The sensitivity analysis 080, fire scenarios were evaluated based on total ODE presented in Appendix B impact and importance of individual fire cutsets or
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR,/F&O # Finding Resolution 5b Impact indicates that the fire growth rate groupings of fire cutsets. More detailed fire modeling was and the burning regime can one of several approaches (e.g., crediting conditional influence the predicted MCR circuit failure probabilities or operator recovery actions) abandonment times given a peak available to remove excessive conservatism and thereby heat release rate. However, to to achieve more realistic risk results. The determination to fully address these parameters in select and apply more detailed fire modeling tools was greater detail would require an based on expert opinion of the expected resulting analysis of individual cabinet improvement. If the chosen approach resulted in a less-enclosures and an assessment of conservative, more-realistic risk, then the risk the fire development and contributions of that scenario diminished and the next ventilation conditions for each most important cutset could be subjected to the process.
| |
| cabinet considered.' Therefore The process also provided for possible plant modification only capability Category I is to reduce the risk for the scenario to acceptable levels considered met. when conservatism was not either readily apparent or easily removed. Although repeated iterations tend to produce diminishing results, this process stopped upon meeting the goal of a FPRA model that provides realistic risk estimates with some reasonable margin to the requirements of RG 1.174.
| |
| BNP-PSA-086 Section 10 A quantitative evaluation of parametric uncertainty for Resolution of this FSS-E3 contains the identified sources of both ODE and LERF was performed as documented in finding provided the (CAT I) uncertainty in the fire modeling EVAL EC 296040, including a State of Knowledge required quantitative FSS-H5 scope. This evaluation was Correlation covering fire ignition frequencies, non- evaluation of (CAT I) limited to a qualitative evaluation suppression probabilities, conditional failure probabilities, uncertainty. This FS-9 of the identified uncertainties. No and fire bins. finding is sufficiently FS-9 statistical representations of the resolved for SR FSS-(CAT I/llI/Ill) uncertainty intervals was present, E3 and SR FSS-H5 to UNC-A2 therefore only Capability be assessed as (CAT 1/11/111) Category I was considered met. meeting CAT II.
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| There is no impact to The heat release rate, the the 5b application.
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| 4-14 shortest distance from the ignition source to the target and the fire
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact diameter are typically considered
| |
| - for statistical representation of uncertainty intervals. The remaining inputs of compartment geometry and ventilation characteristics are obtained from plant drawings are typically not subject to statistical uncertainty analysis.
| |
| BPN-PSA-080 calculation Section Following the methodology in NUREG/CR-6850, BNP- Resolution of this 6 evaluates the impacts of the PSA-080 calculation has been revised and the Multi- finding revised the MCA evaluations. In Section 6 Compartment Analysis does not assume a CCDP of 1.0 analysis for MCA.
| |
| only two MCA scenarios were not for any compartment in the MCA analysis. Compartment This finding is FSS-G6 screened, and required CCDPs were calculated based on actual localized target sufficiently resolved (A ) evaluation. For these two zones, sets for exposing compartments. for SR FSS-G6 to be (A ) the CCDPs were assumed to be assessed as meeting 1 and CLERP was assumed to be CAT IlI/ll. There is no 4-6 .1; therefore no specific impact to the 5b quantitative evaluations were application.
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| performed for these MCAs. As a result Capability Category IlIll1 is not met.
| |
| Existing active components A systematic review of the Level 2 progression for fire Resolution of this LE-E1 identified in the internal events impacts has been performed to the extent required by the finding performed a (NOT MET) models were considered in design of the core damage sequence models and Level 2 systematic review of F-1 component selection and cable model. This review has been documented as the Level 2 FQ1 routing. Quantification was Attachment 14 in the component selection calculation, progression for fire (CAT 1I/Ill11) performed using the existing BNP-PSA-085, Revision 2. impacts. This finding accident progression with no is sufficiently resolved noted changes as related to the for SR LE-E1 to be 4-17 affect of fire scenarios. Existing assessed as meeting modeled operator responses CAT 1I/1I/l1. There is
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact were evaluated for changes due no impact to the 5b to fire affects. ,The MSO application.
| |
| evaluation considered affects of LERF with respect to failure of containment isolation. However, no systematic review of the accident progression to determine if fire affects would impact the existing internal events accident progression was found.
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| Parametric uncertainties that are A quantitative evaluation of parametric uncertainty for Resolution of this QU-E3 associated HLR-DA, HR and IE both ODE and LERF was performed as documented in finding provided the (CAT I) are documented in BNP-01 87. EVAL EC 296040, including a State of Knowledge required quantitative QU-A3 However, the state of knowledge Correlation covering fire ignition frequencies, non- evaluation of (CAT II) correlation was not considered in suppression probabilities, conditional failure probabilities, uncertainty. This UN-l the evaluation of these and fire bins. finding is sufficiently UN-1 uncertainty evaluations, resolved for SR QU-(NOT MET) Correlation should be considered E3 to be assessed as FQ-A4 for fire events such as the fire meeting CAT II and (CAT I/llI/Ill) frequency, applied severity/HRR SR FQ-A4 to be split fractions, non-suppression, assessed as meeting circuit failure probabilities, etc. CAT I/Il/Ill. There is 4-18 no impact to the 5b application.
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| P-P-Al Section 3.2.1 and Attachment 1 of BNP-PSA-083 was revised to add Drywell/Torus, Spent Resolution of this (NOT MET) calculation BNP-PSA-083 have Fuel Pool, and VP1/VP2 to Global Plant Analysis finding revised areas been reviewed to examine the Boundary. The Spent Fuel Pool and Service Water Valve included in the global (CAT I/Il/Ill) process by which the Global Plant Pits were then qualitatively screened, while the plant analysis Analysis Boundary (GPAB) has Drywell/Torus were simply not quantitatively analyzed boundary and the PP-~C2 been defined in the BSEP FPRA. based on no fire being postulated in an inerted criteria for exclusion.
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| (NOT MET) Section 3.2.1 indicates that all atmosphere. This finding is areas that contained any sufficiently resolved equipment or cable credited in the The characterization of equipment as "risk significant" for both SR PP-Al
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding JResolution 5b Impact 5-1 FPRA were included, as well as was removed from the description of the criteria for and SR PP-C2 to be any area would require a plant excluding areas from the GPAB. The distances assessed as meeting shutdown. In addition, any area separating certain buildings of potential interests were CAT I/Il/Il1. There is that is adjacent to an area that added. no impact to the 5b would affect EPRA application.
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| cables/equipment or require a Distance from ABH to DGB is 32'.
| |
| shutdown is said to be included in the GPAB. All of these criteria are Distance from CTPH1 to DGB is 28'.
| |
| in agreement with PP-Al.
| |
| Distance from STORES to RB2 is 30'.
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| However, in Attachment 1, a number of buildings/areas are excluded from the GPAB because they do not affect "risk significant" equipment and they may not require a plant shutdown prior to the assumed threshold of 8 hours (described in Assumption 3.1.1.5). This process is consistent with the guidance provided for the Qualitative Screening Task (task 4) in Section 3.3, but is considered inappropriate for use at the PP stage of the analysis All other areas listed in the table in Attachment 1 should either be confirmed to contain no equipment or cables that are either:
| |
| : 1) credited in the FPRA (i.e., not just risk significant), or
| |
| : 2) capable of adversely impacting
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| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution 5b Impact plant response Additionally, the exclusion basis needs to include additional discussion for the following:
| |
| *Aux. Boiler House (ABH) - State that the closest building of concern is the DG building which is approx ___ ft away and will not be affected by an exposure fire in ABH.
| |
| * CTPH1 - Due to proximity to DG building, discuss-fire exposure potential.
| |
| * Fire house - Any fire alarm panels being affected?
| |
| * STORES - Address exposure to south side of the U2 rector building.
| |
| *VP1, VP2 - Not shown on the BGA boundary drawing.
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| ES-A5 There are a few fire-induced No change will be made to incorporate Finding 5-4 Resolution of this (CAT 1,1II) spurious events that were because further consideration of the listed spurious finding justified the ES-A6 screened, but could in fact either events revealed no additional fire impacts beyond what current approach.
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| (CAT I, II) cause a plant trip (or manual was already identified by the MSO Expert Panel. In There is no impact to ES-B2 shutdown) and impact equipment particular, contrary to Finding 5-4, the events described in the 5b application.
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| (CTII) that is credited for accident Items 1, 2, 4, 5, and 6 neither cause an automatic plant (CT11) mitigation in the FPRA: trip nor require a Tech Spec mandated manual shutdown ES-D1 in less than the 8-hours assumed for treatment as a f ire-(CAT I/Il/Ill) 1) spurious start / injection by induced plant trip.
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| PRM-B9 RCIC. This was screened from (CAT I/l1/l1l) the FPRA and an initiating event Although the event described in Item 3 could either cause because it was assumed that no an automatic plant trip or prompt the Operator to initiate a
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| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions SR/F&O # [Finding -
| |
| JResolution J 5b Impact plant trip would occur. However, a manual scram (depending on the number of individual 5-4 fire-induced RCIC start would control rods that initially scrammed), the RPS scram likely only be caused if the fire signal itself would shortly close the SDV vent and drain damage was significant enough valves, at which point the scenario would most resemble to cause ROIC inoperability. a previous addressed turbine trip.
| |
| Assuming no plant shutdown may be non-conservative. With regard to the suggested possible resolutions:
| |
| For items 1-3, since Section 3.3.2.1 of the MSO report
| |
| : 2) spurious start/Iinjection by (Attachment 4 of Calculation BNP-PSA-085) already HPCI. This was screened from documents significant operator experience for members the FPRA and an initiating event of the MSO Expert Panel, there is little marginal benefit in because it was assumed that no citing additional operator interviews for support.
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| plant trip would occur. However, a For item 4, the only clarification necessary would be to fire-induced HPCI start would note that the item is incorrectly premised on core spray likely only be caused if the fire being able to inject at high RCS pressure.
| |
| damage was significant enough For item 5, the identification in the MSO report of a to cause HPCI inoperability. restricting orifice with a 0.105 inch bore should already be Assuming no plant shutdown may sufficient documentation that the HPCI drain pot line to be non-conservative. the condenser does not constitute a steam flow diversion.
| |
| : 3) MSO item C11 -2e. This MSO For item 6, since an automatic plant trip or manual drains the RPV through the SDV shutdown is required to drop RPV to below that needed vent and drain. The exclusion of for condensate injection, a plot of RPV pressure over time this event from the FPRA is is not needed to invalid this MSO (i.e., spurious based on the fact that the condensate injection with RPV pressure below 500 psig) suppression pool inventory as an initiating event and would add nothing to the depletion is slow and would not evaluation, in the MSO Report (Attachment 4 of reach a low enough level in 24 Calculation BNP-PSA-085), of equipment credited for hours to require a plant trip. post-trip accident mitigation.
| |
| However, it may be nonconservative to assume that there is no chance of a plant trip due to this uncontrolled loss of RCS inventory.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # jFinding ]Resolution j 5b Impact
| |
| : 4) MSO item E21-O01. This MSO describes spurious actuation of the core spray pumps and spurious operation of the injection valves. This event can cause flooding of the main steam lines, which can subsequently cause failure of the turbine-driven RCIC/HPCI pumps (and FW, which is not modeled). The exclusion justification says that high-pressure injection is 'not credited after depressurization',
| |
| so there is no way to model the event. However, if spurious CS pump operation occurred at high RCS pressure and the main steam lines were flooded, HPCI and RCIC should be impacted because there is still potential for crediting their high-pressure injection.
| |
| : 5) MSO item E41 -2w. This MSO describes the unisolated drain of HPCI to the main condenser via spuriously opened AOVs. Two of the three AOVs in series have been locked open, so this scenario only requires one AOV to open (on loss of instrument air or hot short). This event is excluded based on an installed
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact flow-limiting orifice, but there is no technical discussion of the flow limitation to adequately justify why the flowpath is not a valid diversion.
| |
| : 6) MSO item N21-2ai. This MSO describes RPV overfill due to condensate injection once RPV pressure is <500psi. The exclusion justification states that it is unlikely that RCIC/H PCI operation alone would not depressurize the RPV to 500psi in one hour. However, ROIC and HPCI are credited for injection for much longer than 1 hour, at some point RPV pressure may be reduced to allow condensate injection, which could potentially fail HPCI/RCIC.
| |
| Some of the exclusion bases for Section 3.4.3 of BNP-PSA-083 was revised to include Resolution of this the BSEP historical fire events additional discussion of the plant history and corrective finding reviewed the should be strengthened to actions concerning fires related to the heater drain pumps fire events in question IGN-A4 support the conclusion that the (Items #1, #2, #5, and #7). The appropriate exclusion of and revised the (CAT 1,II) use of generic ignition frequency Item #3 as being outside the GPAB was confirmed. The calculation, as IGN-B4 data is appropriate: appropriate exclusion of Item #4 and Item #6 as being not necessary, to (CAT VIIl/Ill) potential challenging was confirmed. Some further document the basis
| |
| : 1) FR 88-006: A heater drain clarification of the documentation for Items #3, #4, and #6 for considering them pump ignited and required 3 CO2 was considered but judged unnecessary at this time. as being not 5-8 extinguishers at power. potentially Approximately 2 quarts of oil were challenging. There is burned. This appears to be no impact to the 5b potentially challenging, application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # fFinding jResolution 5b Impact
| |
| : 2) FR 90-002: A heater drain pump ignited at power and required 'several' extinguishers.
| |
| Fire was fueled by pump oil, caused a fire alarm, and resulted in -$64k worth of damage. This appears to be potentially challenging.
| |
| : 3) FR 94-007: A OWOD pump ignited at power and required offsite fire department response.
| |
| IF this was not dismissed in PP, this could be a potentially challenging fire.
| |
| : 4) ACR 94-01 488: A fire in a Rad Waste control room panel at power required a fire extinguisher. Fire caused loss of SFPC, which appears to be potentially challenging.
| |
| : 5) ACR 97-1136: A heater drain pump ignited at power and was secured to extinguish the fire in response to the fire alarm. This appears to be potentially challenging.
| |
| : 6) ACR 98-651 : A cable fire started in a manhole at power due to water intrusion and
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # JFinding Resolution 5b Impact corrosion. Fire was self-extinguishing, but cable damage was reported.
| |
| : 7) NCR 24699: A heater drain pump ignited an oil fire, which caused a fire alarm and required 002 extinguishers while at power. A condensate system transient resulted and an unusual event was declared due to a duration of >10 minutes. This appears potentially challenging.
| |
| There is no record of a review A systematic review for modeling inconsistencies Resolution of this being performed to confirm that finding performed a associated with fire impacts was performed and has the FPRA modeling is consistent systematic review for been documented in various sections of BNP-PSA-085, from event sequence to system modeling with conclusions consisting of statements of acceptability model or that with operational inconsistencies QU-D2 or descriptions of the required changes.
| |
| (NOT MET) characteristics. Since the FPRA associated with fire Section 3.3.1.4 of BNP-PSA-085 describes the review of model is largely based on the impacts. This finding QU-F3 the PRA Internal Events model Accident Sequence (CAT I) internal events model, this is is sufficiently resolved Notebook (i.e. BNP-PSA-029) and Level 2 Accident assumed to be a relatively for SR QU-D2, SR F'Q-E1 Sequence Notebook (i.e. BNP-PSA-049). These PRA insignificant source of potential EQ-El, and SR FQ-(NOT MET) notebooks address the plant response and the event model inaccuracy. However, a F1 each to be EQ-Fl trees developed for that response. The review determined review does need to be assessed as meeting that the EPRA will be maintained as part of the Internal (NOT MET) performed to confirm that fire- CAT I/ll/l1l and for SR Events PRA Model of Record and concluded that the use specific modeling considerations QU-F3 to be of the same PRA models as for the internal events have not created any assessed as meeting 5-13 sequence quantification ensures that inconsistencies between CAT IlI/ll. There is no interdependencies are modeled consistently and sequence and system modeling, impact to the 5b appropriately. Because the BSEP PRA uses functional or between the EPRA model and tops for the event trees, and functions are modeled by application.
| |
| actual plant operational practices. initiating events and by system models, a detailed sequence by sequence review is not required and would
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding [Resolution 5b Impact provide no benefit. The fire events cannot change the modeled functions, and the conclusion is always that the accident sequence logic is adequate because the functions are always the same: reactivity control, RPV integrity, inventory control, pressure control, decay heat removal, and containment integrity.
| |
| Section 3.3.1.4 of BNP-PSA-085 also describes the review of applicable initiating events (i.e., BNP-PSA-032) and the relevance to the fire model (i.e., BNP-PSA-085, Attachment 8). Consideration of possible additional initiating events that might be unique to the FPRA is documented in Table 3-2 of BNP-PSA-085.
| |
| Attachments 3 and 4 of BNP-PSA-085 document the review and disposition of various fire-induced MSOs postulated by plant and industry personnel to have potential impact on mitigation functions and systems. This review resulted in certain FPRA model changes as documented in Attachments 9 and 12 of BNP-PSA-085 and included the creation of a simplified bypass event tree for a main steam isolation valve (MSIV) MSO, as described BNP-PSA-085, Section 3.3.1.4 and Attachment 13.
| |
| For the Level 2 review, a detailed review of the containment isolation is performed in BNP-PSA-085, Attachment 6. Subsequently the review of the fire impact on the Level 2 accident sequences and phenomenological events was performed as documented in Attachment 13 of BNP-PSA-085.
| |
| QU-F3 A review of the cutset review Non-significant cutsets were reviewed and the results are Resolution of this (A ) documentation indicates that the documented in Attachment 39 of BNP-PSA-080 (i.e., finding reviewed non-(A ) vast majority (if not all) of the Change Package BNP-0235). significant cutsets.
| |
| QU-D5 reviewed cutsets are from This finding is (NOT MET) significant scenarios, almost ______________________sufficiently resolved
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact EQ-El exclusively with CCDPs of 1.0. for SR QU-D5, SR (NOT MET) Many of these CCDP cutsets EQ-El, and SR FQ-E-l have only a single cutset (other Fl each to be (NOT MET) applicable cutsets are truncated). assessed as meeting At the current stage of the BSEP CAT I/Il/Ill and for SR EPRA development, this is not an QU-F3 to be 5-4 unreasonable characteristic of the assessed as meeting cutset reviews. CAT IlI/ll. There is no impact to the 5b However, a lack of review of non- application.
| |
| significant cutsets precludes
| |
| ___________meeting this SR.
| |
| Although a limited identification of The correctness and reasonableness of the FPRA Resolution of this fire CDF contributors has been modeling were reviewed based on detailed cutset reviews finding documented performed, the types of for individual scenarios, such as that documented the review of FPRA QU-E2 contributors is limited and there is in Attachment 39 of BNP-PSA-080. The results of the results and the (CAT 1/1l/I1l) little or no discussion of the risk Fire PRA have been reviewed and top contributors by identification of risk QU-F3 insights gained from the ignition source, including transient and fixed, and insights. This finding (CAT I) contributor identification. compartment have been identified and listed in Section is sufficiently resolved Q-63.4 of Attachment 38 to BNP-PSA-080, by cutset for SR QU-D7, SR QU-D6I For example, the following probability, and the top contributors to both ODE and EQ-El, and SR FQ-(CATuin I)db nihtu, LR.Te ikcnrbtosadrs mprac vns 1ec ob QU-D7 contribtion oulbee idnsgtiful, LERd Theairsk conributosesd anddsrisk imoanc d eventso El3 assedach tomeetn (NOT MET) anoaentbe ietfe: adfAtailurest wer asBN-Seed , asnecrbd in Sctuerniong33 ssessed/I asdmeeting FQ-Eof-Attachmentn38atoidentPsAq8encsand includedorankingseCATiI/Il/Illand forQSR (NOT MET) - risk significant operator actions -Fire Compartments F3 to each be EQ-Fl performed inside the main control -Fire Scenarios assessed as meeting (NOT MET) room -Fire Accident Sequences CAT IlIll1. There is no
| |
| - risk significant operator actions -Containment Failure Types (i.e., LERE only) impact to the 5b performed outside the main -Operator Actions application.
| |
| 5-15 control room -Fire Induced Equipment Eailure Modes
| |
| - contribution to fire CDF from -admCmoetFiue transient ignition sources
| |
| - contribution to fire ODE from -Systems
| |
| | |
| Table 3. BSEP Fire Peer Review Findings & Observations Resolutions SR/F&O # {Finding -
| |
| Resolution 5b Impact fixed ignition sources -Component Type Failures
| |
| - significant spurious actuation For each of these categories other than Containment events Failure Types, the top contributors were ranked for both
| |
| - significant random failure events CDF and LERF according to the percent contribution to (i.e. non-fire), including common risk. For Containment Failure Types, the assessment only cause failures considered the contributions to LERF. The importance
| |
| - the REDUCTION in ignition ranking results in each of these ranking categories are frequency contribution to fire CDF generally used in addressing which portions of the FPRA due to the extensive use of the model need further refinement.
| |
| conditional plant trip probabilities Insights from importance analysis were used to review the correctness and reasonableness of the FPRA modeling Additionally, the importance of by comparing the results against what is normally components and basic events understood about plant response. Attachment 38 of BNP-were not reviewed to determine PSA-080 provided the following insights for reviewing the that they make logical sense (QU- correctness and reasonableness of the FPRA model:
| |
| D7). -The MCR and cable spreading rooms dominate risk contributions from Fire Compartments;
| |
| -Scenarios involving fixed ignition sources, rather than transient combustibles, are major contributors;
| |
| -All transient and station blackout (SBO) sequences for fire result in either loss of makeup events or loss of decay heat removal events that result in a loss of makeup;
| |
| -With emergency power blackout associated with the dominant cause of core damage, failures of fail-safe containment isolation valves may not contribute as much to LERF as presented;
| |
| -Control room abandonment for habitability is one of the more important operator actions;
| |
| -Lack of a specific method for evaluating fire-induced instrument faults is evident in the results;
| |
| -Random component failure rankings show test and maintenance unavailability is important to fire risk;
| |
| -The plant system that contributes most to fire risk is the
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b impact AC power system, with AC breakers as contributing components.
| |
| With respect to identifying the The correctness and reasonableness of the FPRA Resolution of this contributors to fire LERF, the modeling were reviewed based on detailed cutset reviews finding documented following contributors are for individual scenarios, such as that documented the review of EPRA considered: in Attachment 39 of BNP-PSA-080. The results of the results and the Fire PRA have been reviewed and top contributors by identification of risk
| |
| - contributions from fire scenarios, ignition source, including transient and fixed, and insights. This finding LE-Fl MCA abandonment, and the compartment have been identified and listed in Section is sufficiently resolved (NOT MET) multi-compartment analysis. 3.4 of Attachment 38 to BNP-PSA-080, by cutset for SR LE-F2, SR FQ-
| |
| - compartments with >1% fire probability, and the top contributors to both CDF and El, SR EQ-Fl and SR LE-F2 LEFLERF. The risk contributors and risk importance events UNC-A1 each to be (NT ET -ignition sources with >1% LERE and failures were assessed, as described in Section 3.3 assessed as meeting I-E-G3 of Attachment 38 to BNP-PSA-80, and included rankings CAT I/Il/Ill and for SR (CAT I) No identification of plant damage in the following categories: LE-F1 and SR LE-G3 UNC-A1 states or containment failure -Fire Compartments each to be assessed (NOT MET) modes was identified, which is -Fire Scenarios as meeting CAT IlI/ll.
| |
| EQ-El required for CCI. To meet CClI, -Fire Accident Sequences There is no impact to (NOT MET) additional identification of -Containment Failure Types (i.e., LERE only) the 5b application.
| |
| EQ-El significant fire LERE contributors -Operator Actions (NOT MET) is required, as discussed in the -Fire Induced Equipment Failure Modes SR. -Random Component Failures Within the scope of fire LERF -Systems 5-16 contributors that have been -Component Type Failures identified, it is not apparent that a For each of these categories other than Containment review for 'reasonableness' has Failure Types, the top contributors were ranked for both been performed. CDF and LERF according to the percent contribution to risk. For Containment Failure Types, the assessment only For example, 98.1% of Unit 2 fire considered the contributions to LERF. The importance LERE is due to fires in the Unit 2 ranking results in each of these ranking categories are
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # Finding ]Resolution 5b Impact main control room. Although this generally used in addressing which portions of the FPRA is identified in table 11-2 of BNP- model need further refinement.
| |
| PSA-080, there is no discussion Insights from importance analysis were used to review the of this considerable contribution correctness and reasonableness of the EPRA modeling including whether or not it is by comparing the results against what is normally considered reasonable. Notably, understood about plant response. Attachment 38 of BNP-the Unit 1 MCR contributes -60% PSA-080 provided the following insights for reviewing the of Unit 1 fire LERF, and no correctness and reasonableness of the FPRA model:
| |
| discussion of this asymmetry is -The MCR and cable spreading rooms dominate risk provided contributions from Fire Compartments;
| |
| -Scenarios involving fixed ignition sources, rather than transient combustibles, are major contributors;
| |
| -All transient and station blackout (S60) sequences for fire result in either loss of makeup events or loss of decay heat removal events that result in a loss of makeup;
| |
| -With emergency power blackout associated with the dominant cause of core damage, failures of fail-safe containment isolation valves may not contribute as much to LERF as presented;
| |
| -Control room abandonment for habitability is one of the more important operator actions;
| |
| -Lack of a specific method for evaluating fire-induced instrument faults is evident in the results;
| |
| -Random component failure rankings show test and maintenance unavailability is important to fire risk;
| |
| -The plant system that contributes most to fire risk is the AC power system, with AC breakers as contributing components.
| |
| LE-G2 Assumptions for the quantification A quantitative evaluation of parametric uncertainty for Resolution of this (NOT MET) task are documented in Section both CDF and LERF was performed as documented in finding documented LE-F3 3.3 of BNP-PSA-080. General EVAL EC 296040, including a State of Knowledge the review of sources (NOT MET) sources of uncertainty are Correlation covering fire ignition frequencies, non- of uncertainty for LE-G4 discussed in Section 8.4. These suppression probabilities, conditional failure probabilities, LERF. This finding is
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SRIF&O # [Finding Resolution j 5b Impact (NOT MET) sources include: and fire bins. sufficiently resolved UNC-A1 for SR LE-F3, SR LE-(NOT MET) - ignition frequencies As evaluated in BNP-PSA-080, Attachment 38, the most G2, SR LE-G4. SR UNC-A2
| |
| - HRRs significant area of epistemic uncertainty with regard to UNC-A1, SR EQ-El, (CAT I/Il/Ill)
| |
| - target selection LERE is in circuit analysis, specifically as related to and SR EQ-El each
| |
| - damage time spurious operation of containment isolation valves. to be assessed as EQ-El - time to HGL Lacking specific guidance in NUREG/CR-6850 for the meeting CAT I/Il/Ill.
| |
| (NOT MET) - fire effects treatment of low voltage instrumentation loops and the There is no impact to EQ-Fl - suppression grounding or clearing hot shorts in DC circuits, these the 5b application.
| |
| (NOT MET) - circuit analysis events were assigned a value of 1.0 in the BSEP FPRA.
| |
| -HRA The resulting cutsets are dominated by signal failures
| |
| - quantification (including tools) causing valve spurious operations or primary containment 5-18 isolation valves (PCI Vs) remaining spuriously open, even These sources of uncertainty are though their design is to fail safe in the closed/isolated valid in the fire LERF and fire position. A more realistic assessment of these affects ODE quantifications, but there are would greatly reduce LERF.
| |
| no additional sources of uncertainty that are applicable to The sources of aleatory uncertainty were evaluated for the fire LERE calculation. Change LERF, and a detailed results analysis was performed for package BNP-O0187 provides fire LERF as documented as Attachment 38 to the BNP-PSA-CDF importance measures and a 080 and supplemented in EVAL EC 292418. This analysis statistical analysis of fire CDE includes evaluation of parametric uncertainty for a uncertainty, but does not address combined LERF solution and various importance fire LERF. evaluations for the same solution, including the State of Knowledge Correlation. In particular, parametric uncertainty was evaluated for the following:
| |
| (a) Fire scenario event frequencies (i.e., initiators),
| |
| (b) Component failure probabilities (i.e., random faults and hot short probabilities),
| |
| (c) Component maintenance unavailability, (d) Human error probabilities, (e) Common cause failures, and (f) Recovery Actions (i.e., main control room abandonment from environmental causes)
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The types of fire data reviewed for SOKO included: fire ignition frequency, non-detection probabilities, non-suppression probabilities, heat release rate severity factor/split fraction, and circuit failure probabilities.
| |
| The top contributors to LERE were ranked for: fire compartments, fire scenarios, accident sequence types, containment failure types, operator actions, fire-induced equipment failure modes, component types, and failure
| |
| _________________________groupings.___________
| |
| There is no definition established A discussion of "significance" in terms of the definitions Resolution of this for 'significance' related to basic described in ASME/ANS-Ra-Sa-2009 Section 1-2 has finding added a LE-G6 events, cutsets, accident been added to Section 8 of the Quantification Calculation discussion of (NOT MET) sequences, or any other facets of (i.e., BNP-PSA-080). significance to QU-F6 the fire PRA results. documentation. This (NOT MET) finding is sufficiently FQ-F1 resolved for SR LE-(NOT MET) G6, SR QU-F6, and SR FQ-F1 each to be assessed as meeting 5-19 CAT I/Il/Ill. There is no impact to the 5b application.
| |
| Thedefciet sb-equremntsofAttachment 13 was added to BNP-PSA-085, Revision 2, Rslto fti this SR are detailed below, to address items A, B, D, E, F, and part (i.e., equipment, finding added LE-G2 A) No documentation was containment failure modes and phenomena) of C. documentation to (NOT MET) provided of plant damage states / address the FQ1 attributes, although this can be The remainder of item C (i.e., fire-specific human actions requirements. This FQ-F1 considered covered by general considered in the fire LERF sequence development) was finding is sufficiently (NOT MET) references to the internal events addressed in Section 4.2.3 and Table 5.1 of BNP-PSA- resolved for both SR PRA model. 084, Revision 2. LE-G2 and SR EQ-Fl 5-20 B) There is no documentation of F&O 5-18 sufficiently addresses item H (i.e., LERF- to be assessed as how accident sequences were related uncertainty). Resolution was completed as part of meeting CAT I/Il/lll.
| |
| binned into plant damage states, F&O 5-18. There is no impact to but since the fire LERF model is the 5b application.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # ]Finding ]Resolution 5b Impact based on the internal events LERF model, references to the internal events PRA can account for this.
| |
| C) There SHOULD be discussion of the fire-specific human actions and equipment considered in the fire LERF sequence development. Containment failure modes and phenomena could be referenced to the internal events documentation D) There is no discussion of fire-specific factors influencing containment challenges and containment capability.
| |
| E) Containment capacity analysis could be covered by a reference to the internal events LERF model. No fire-specific impacts are expected.
| |
| F) A discussion of fire-specific impacts on the accident sequences identified in the containment event trees should be provided.
| |
| H) The model integration process is described in Section 4.9 of BNP-PSA-080. There are no fire LERF-related uncertainty (F&Q 5-
| |
| : 18) or sensitivity analyses provided.
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution J 5b Impact There is evidence that the Evidence of the collective review of electrical coordination Resolution of this existing electrical coordination with supporting analysis for breakers, power supplies and finding reviewed and analysis was reviewed and cables was documented in BNP-PSA-080 as: documented electrical refined (e.g. BNP-O0157). Specific Attachment 13 (i.e., Change Package BNP-0157) coordination with documentation should be Attachment 36 (i.e., Change Package BNP-0218) supporting analysis provided of this review. There is Attachment 37 (i.e., Change Package BNP-0224) for breakers, power no evidence that power supplies Attachment 41 (i.e., Change Package BNP-021 5) supplies and cables.
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| credited in the fire PRA were Attachment 42 (i.e., Change Package BNP-0217) This finding is reviewed to confirm that they Attachment 43 (i.e., Change Package BNP-0223) sufficiently resolved were addressed by existing for SR CS-B 1 to be overcurrent calculations. assessed at meeting Attachment 36 of BNP-PSA-080 identifies two raceways CAT IlIll1 and SR CS-that could not be routed on drawings. However, C4 to be assessed as Section 5.d of Attachment 36 describes how a reasonable meeting CAT I/Il/Ill.
| |
| CS-B1 approximation for the fire zones that the raceways There is no impact to (CAT I) traverse was possible based on the raceways adjacent to the 5b application.
| |
| CS-C4 the raceway in question, the cable start and end (NOT MET) equipment, and general plant layout knowledge.
| |
| Attachment 37 of BNP-PSA-080 identifies three other 6-1 raceways that could not be routed. The raceways that are missing information are located in the Unit 2 electrical equipment room (i~e., Control Building 49'). The raceway for 2-2A-120V involves a cable running between two panels in adjacent rows in the Unit 2 electrical equipment room (i.e., Control Building 49'). The two raceways for 2-2D-120V involve a cable running from a panel in the Unit 2 electrical equipment room (i.e., Control Building 49') to a panel in the Unit 2 cable spreading room (i.e., Control Building 23'), which is where the two other raceways are known to be located. These three raceways are identified as a source of uncertainty in Section 3.6.6 of BNP-PSA-080, and the risk associated with their assumed failure is qlualitatively addressed as a
| |
| | |
| Table 3. BSEP - Fire Peer Review Findings &Observations Resolutions SR/F&O # Finding Resolution 5b Impact non-conservative assumption (i.e., Section 3.1.3.44) that is likely mitigated in the HGL scenario by other failures for the respective power supplies.
| |
| All panels modeled in the FPRA were included within the scope of the breaker coordination study, as described in Attachment 42 of BNP-PSA-080.
| |
| Passive fire barriers with a fire Section 11.5.4 of NUREG/CR-6850 requires postulating Resolution of this resistance rating are credited in the failure of only one fire barrier, and selecting the worst finding justified the the multicompartment analysis. case value for those applicable to the Fire Compartment current approach.
| |
| FSS-G4 The failure rates used are those is conservative. There is no guidance for summing the This finding is (CAT I) prescribed in NUREG 6850, probabilities for individual elements over an entire wall to sufficiently resolved however, the worst case value for get probability of wall failure. Walkdowns were performed for SR FSS-G4 to be failure probability of the barrier is to gather the targets and barriers between the exposing assessed as meeting 6-4 used. and exposed compartments. The worst case barrier CAT II. There is no failure probability was applied to all local targets between impact to the 5b two adjacent compartments. The results of this analysis application.
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| is included in Attachment 7 of BNP-PSA-080.
| |
| Screening methodology is As described in Attachment 7 to BNP-PSA-080, plant Resolution of this provided in BNP-PSA-080, walkdowns were performed to identify targets in the revised the calculation Section 6.0. exposed compartments near the barriers separating the to include the exposing and exposed compartments. The localized localized damage in FSSG2 However: the MCA screening did damage in the adjacent compartment near barriers for all the adjacent (CAT I/Il/Ill) not consider the impact of compartments that screened out and for compartments compartment near possible localized effect (e.g., where MCA was performed but did not achieve a HGL in barriers. There is no damage to equipment) near the combined compartments was included. The localized impact to the 5b 6-5 penetrations and barriers, targets of the adjacent compartment were added to the application.
| |
| HGL evaluation for the exposed compartment.
| |
| In addition, a screening value was used without justification and the cumulative risk for the screened scenarios was not evaluated.
| |
| | |
| Tab&O#lenin 3.RE iePe eiwFnig bervautions Resolution SRIF&O # Finding Resolution 5b Impact Conditional failure probabilities No non-conservative application of conditional failure Resolution of this were assigned to selected cables probabilities has been identified for an off-scheme cable. finding justified the per the methodology identified in current approach.
| |
| BNP-01 37, which is based on the For safe shutdown, any failure of an associated circuit This finding is Chapter 10 tables in NUREG also fails the main component. This is conservative, in sufficiently resolved 6850. However the BSEP that (typically) only one or two of an associated circuit's for SR CF-Al to be methodology for determining the cables actually affect the primary component. When assessed as CAT component level spurious applied to the Fire PRA, this method of including IlIll1. There is no operation probability, as identified associated circuits created far too many false failures, impact to the 5b in BNP-PSA-080 Section 4.3.4 and therefore associated circuits are not always linked to application.
| |
| CF-Al and 4.6.1.2.4, is to use the worst the primary component as shown in FSSPMD. In almost (CAT I) case spurious operation all cases, the associated circuits are modeled separately probability of all affected cables as primary components in the fire PRA fault tree. In this without regard as to whether the manner, cable damage to the associated circuit is 6-7 cables in question are primary captured within the fault tree, and will cause cascading scheme or off-scheme cables. failures based on the model. In addition, key interlocks Per FAQ 08-047, off scheme that can have an impact on the Fire PRA are included in cables and cables with alternate the model. Therefore it can be determined that off-source breakers must be scheme cables are included.
| |
| identified and, when combined with on-scheme cables, an Additionally, many times, although they are included, the exclusive OR must be used. failure probability may be 1.0, and appear to be Spurious events of high unanalyzed. In assigning the fault probabilities for importance that had spurious Brunswick, specific basic events were identified by PRA.
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| | |
| Table 3. BSEP - Fire Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact operabilities applied were Fault probabilities were assigned to the on-scheme reviewed and found to have no cables that could affect the basic event of concern. The off-scheme cables, therefore Cat I values assigned represented the best-estimate as shown is considered met. in the tables in Chapter 10 of NUREG/CR-6850. These fault probabilities were in general, only applied to control circuits. A loss of power that results in the failure of a basic event could occur due to a short to ground, and since the fault probabilities provided in NUREG/CR-6850 only apply to hot shorts, a probability of 1.0 would be assigned. Similarly, instrumentation cables are assumed to fail with a probability of 1.0 since they have not been specifically tested. However, since they can fail either high or low, a split fraction may still be applied to the functional response to the cable fault. Since many of the associated circuits are tied to instrumentation, not performing a fault probability analysis on such circuits has no impact on the PRA results since the failure would be an assumed value of 1.0, and no advantage would be gained.
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The wind PRA model should External exposed components are failed through the use This issue has been include only those SSCs of a flag file. Components that reside in turbine building assessed and addressed by the fragilities are assumed protected from wind hazard to the same changes will be analysts, and trim the rest from extent as the loss of offsite power fragility. Most of applied to the model.
| |
| WP-3 PRA. components in turbine building are lost with the loss of There is no impact to (NOT MET) offsite power which is much more fragile than the turbine the 5b application.
| |
| building. Exposed components have been assessed and WPR-A3-0 1 dispositioned accordingly. Components fed by offsite power (circulating water pumps and startup transformers) are assumed as rugged as offsite power source that powers them. The component Basic Events failed by wind events have been identified.
| |
| From the current documentation, For long duration high wind events (hurricanes) operators This issue has been it is not apparent why only 3 will be stationed in the turbine building to perform certain assessed and HEPs were selected for multiplier HRA events (outlined in procedure 0AOP-13). For short changes will be effects from high winds, even duration events (high straight line and tornado winds) the applied to the model.
| |
| WPR-A5 though Attachment 13 shows event will end before the human action is needed. Since There is no impact to (NOT MET) many HEPs for which multipliers storms are not uncommon events, the performance the 5b application.
| |
| WPR-A8 would apply. shaping factors following the event should be minimally (NOT MET) impacted. The only actions that should be prohibited are events outside the control room, turbine building, reactor WPR-A5-O01 building or diesel building that occur within the first hour.
| |
| The model has been updated to fail short term external actions since they may not be viable and removes multipliers on the HEPs.
| |
| WPR-A5 There are two potential errors in Upon further evaluation it was determined that short term This issue has been (NOT MET) utilizing the multiplier criteria in external actions should be failed since they may not be assessed and WPR-A8 Table 4. First, ex-control room viable. changes will be (NOT MET) actions that do not traverse applied to the model.
| |
| through areas impacted by winds There is no impact to and performed in areas not the 5b application.
| |
| WPR-A5-02 impacted by winds do not need to
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact be considered at a higher failure probability than in-control room actions. This is potentially conservative.
| |
| Second, the criteria in Table 4 do not recognize the possibility that some ex-control room actions may have a guaranteed failure (e.g., an action that requires going outside in a severe hurricane). The maximum increase in the HEPs in Table 4 is a factor of 30. This is potentially non-conservative.
| |
| With high failure probabilities, The product of the high wind initiator and the given No further action there is the potential for success component's fragility are applied as direct failures to all of required. Therefore, branches of the event tree to be the applicable SSCs (product of failure frequency and there is no impact to overestimated. The quantification initiator frequency is very small) as such the assumption the 5b application.
| |
| WPR-A9 engine applies a min-cut upper of small probabilities maintains the validity of the min-cut (NOT MET) bound approximation of the point upper bound estimate to provide accurate results. As estimate, at higher wind intervals such, there is very little impact from not modeling the WPR-A9-01 the fragility values approach 1.0 complimentary success state.
| |
| and mmn-cut upper bound estimate is not sufficient to provide accurate results.
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact If system recoveries (e.g., service Upon further evaluation it was determined that recovery This issue has been water recoveries) are credited in actions that are not feasible due to equipment damage assessed and the WPR-A1 1 the model, then their potential to should be failed. The rule recovery file and flag file was changes will be (CAT I) be impacted by the high winds updated to reflect the changes made to this HRA applied to the model.
| |
| conditions needs to be evaluated, assessment. There is no impact to WPR-A1 1-01 the 5b application.
| |
| The requirement is to assess the Accident sequences have been reviewed and impacts of There is no impact to accident sequences. While the assumptions have been considered. This is a the 5b application.
| |
| ability to assess the sequences documentation issue.
| |
| WPR-B1 has been developed, there is no (CAT I) documentation of the high winds (CAT I) quantification. It is also beneficial WPR-B -01 to review the results and consider the impact of any conservative assumptions that drive the results.
| |
| The requirement is to account for Error factors have been estimated from the range on the This issue has been the uncertainties in each of the hazard curves and an error factor of 10 has been assessed and inputs and for all important assumed for other basic events. The error factors have changes will be WPR-B2 dependencies and correlations. been added to the .RR file. This Finding pertains to applied to the model.
| |
| (CAT I) This has not been performed. documentation of this issue. There is no impact to the 5b application.
| |
| WPR-B2-01
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SRIF&O # Finding Resolution 5b Impact The requirement is to document This Finding pertains to a documentation issue. Additional the specific adaptations to the documentation internal events PRA to produce delineating the specific WPR-C2 the high-wind PRA. This has not adaptations to the (NOT MET) been performed. internal events PRA model will not impact WPR-C2-01 the insights and results used to support the 5b application.
| |
| Sources of uncertainty and This Finding pertains to documentation of sources of Further documentation assumptions must be collected to uncertainties and assumptions in the model. of this issue does not meet the SR. It is also beneficial impact the insights (NOT-MET to characterize and assess the and results used to (OME) impact the sources of uncertainty support the 5b WPR-C3-01 may have on the model and application.
| |
| results.
| |
| XFPR-A3 The requirement is to ENSURE Evaluation of the potential external flooding impact in Resolution of this (NOT MET) the PRA models reflect external operator actions, how operator actions that are not finding evaluated and XFPR-A5 flood-caused failures. To provide credited were added to the PRA model, and a fragility documented potential (NOT MET) this assurance documentation is analysis for equipment was addressed. Two human external flooding XFPR-A8 needed for the systematic review reliability actions were not considered feasible: OPER- impact in operator (NOT MET) for potential impacts of external SWDISCHX and OPER-RESOSP. actions. There is no XFPR-A10 flooding. impact to the 5b (NOT MET) application.
| |
| XFPR-A3-01__________
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding Resolution 5b Impact The likelihood of common-cause Potential for common cause failure due to large debris Resolution of this failures are an important related to external flooding was incorporated into the finding documented dependency. For example, the documentation and analysis. However, the fact that the the potential for XFPR-A7 clogging of intake structures and plant will not be at 100% power during severe weather common cause failure (NOT MET) other flow paths by debris related conditions, would minimize the amount of debris being due to large debris to the flooding should be sucked into the canal and subsequently to the intake related to external XFPR-A7-01 considered, and the walkdown is structure as the circulating water pumps would not be at flooding. There is no a means to ensure that this issue full power since the unit would be in shutdown mode. impact to the 5b has been evaluated properly. application.
| |
| If system recoveries (e.g., service System recoveries consideration and their potential to be Resolution of this water recoveries) are credited in impacted by the external flooding conditions were finding evaluated the model, then their potential to evaluated. The BSEP model does not consider any repair system recoveries XFPR-A1 1 be impacted by the external work after component failure. All failed systems are not potentially impacted (CAT I) flooding conditions needs to be considered recoverable during the transient, including by external flooding.
| |
| evaluated. loss of off-site power (LOOP). There is no impact to XFPR-A1 1-01 the 5b application.
| |
| The requirement is to document Modifications to the BSEP model include: A flag (FL- Resolution of this the specific adaptations to the EXTELOOD) to enable the external flooding model and finding documented internal events PRA to produce two new initiating events %EXTFLI (under gate the specific the external flood PRA, and to ACPTEEXTFLOOD-L) and %EXTFL_2 (under gate adaptations to the document the final results as well ACPTEEXTFLOOD-S), failing the LOOP recovery, internal events PRA to XFPR-C2 as selected intermediate results. %EXTFL_2 is a below design basis initiating event that produce the external (NOT MET) This has not been performed. models a 20 ft still water flood event and has an initiating flood PRA. There is event frequency of 7.4E-04/yr. %EXTFL_1 is a beyond no impact to the 5b XFPR-C2-01 design basis initiating event that models a 23 ft still water application.
| |
| flood event and has an initiating event frequency of 5E-05/yr. Both events fail the switchyard, the electric and diesel firewater pumps, and the circulating water pumps, except that %EXTFL_1 also fails the emergency diesel generators. However, as there is a very large degree of
| |
| | |
| Table 4. BSEP High Winds and External Flooding Peer Review Findings & Observations Resolutions SR/F&O # Finding ]Resolution 5b Impact uncertainty related to the frequency and duration of the external flooding event associated with the 23 ft still water flood (%EXTFL_1), this event is set to '0.0' in the model, while %EXTFL_2 is set to its nominal value. OPER-RESOSP and OPER-SWDISCHX are failed at both flooding scenarios, and therefore are set to 'TRUE' in the flag file. An error factor (EF) of 10 was added to each flooding initiating event to compensate for the uncertainty of their values. Quantification documentation was expanded to also include results and insights based on the cutset review.
| |
| | |
| BSEP 15-0101 Enclosure 3 Marked-up Technical Specification Pages - Unit 1
| |
| | |
| Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between the monitored Once within core keff and the predicted core keff is within + 1% Ak/k. 24 hours after reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 14 OONRI\I-MDT
| |
| '7 oprthecnse-dn MOE !~-H B
| |
| lIn accordance with the Surveillance Frequency Control Program
| |
| /
| |
| II Brunswick Unit 1 3.1-6 BrunwickUni 1 31-6Amendment No. 2-82
| |
| | |
| Control Rod OPERABILITY 3.1.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 2A. he'-rs L-.4 SR 3.1.3.2 ---------- NOTE-------.............
| |
| Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER greater than the L.PSP of the RWM.
| |
| ... is 4
| |
| 7' Insert each withdrawn control rod at least one notcl; m
| |
| SR 3.1.3.3 Verify each control rod scram time from fully In accordance with to notch position 06 is < 7 seconds. SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
| |
| In accordance with the Surveillance Frequency Control Program IJ Brunswick Unit 1 3.1-10 Bruswik Uit 3.-10Amendment No. 2-50
| |
| | |
| Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify, for a representative sample, each tested ,200-4...
| |
| control rod scram time is within the limits of c-wmIat-ive Table 3.1.4-1 with reactor steam dome pressure ..... t""n ""
| |
| > 800 psig. MOat SR 3.1.4.3 Verify each affected control rod scram time is wi in Prior to declaring the limits of Table 3.1.4-1 with any ratrsecontrol rod dome pressure. OPERABLE after work on control rod or CRD System that
| |
| * could affect scram time SR 3.1.4.4 Verify each affected control rod sc am time is within Prior to exceeding the limits of Table 3.1.4-1 with r trsemdome 40% RTP after fuel presure pig.movement__ 00 within the affected core cell in acordncewiththeAND Control Program Prior to exceeding 40% RTP after work on control rod or CR0 System that could affect scram time
| |
| ........*...L.Unit 1 i.*l *.lu u.,,,vv n*,u*
| |
| 3.1-13 Bfuswik No. 2-38 I Uit 3.-13Amendment
| |
| | |
| Control Rod Scram Accumulators 3.1.5 ACTIONS (continued) _________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control rod C.1 Verify all control rods Immediately upon scram accumulators associated with inoperable discovery of inoperable with reactor accumulators are fully charging water steam dome pressure inserted, header pressure
| |
| < 950 psig. < 940 psig AND C.2 Declare the associated 1 hour control rod inoperable.
| |
| D. Required Action B.1 or C.1 D.1-------NOTE- --
| |
| and associated Completion Not applicable if all Time not met. inoperable control rod scram accumulators are associated with fully inserted control rods.
| |
| Manually scram the reactor. Immediately SURVEILLANCEREQUIREMENTS _______
| |
| SU RVEI LLANCE FREQ UENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure is 7-d*y
| |
| Ž 940 psig. '
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3,1-17 BrunwickUnit13.-17Amendment No.
| |
| | |
| Rod Pattern Control 3.1.6 ACTIONS (continued)_________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1------------NOTE-----
| |
| control rods not in Control rod may be compliance with BPWS. bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1.
| |
| Suspend withdraw'al of Immediately control rods.
| |
| AND B.2 Manually scram the reactor. 1 hour SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE I FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with BPWS. hc'-r~i~s SIn accordance with the Surveillance Frequency Control Program I tl Brunswick Unit 1 3.1-19 BrunwickUnitI3.-19Amendment No. 24O*
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.I Restore SLC subsystem to 7 days inoperable. OPERABLE status.
| |
| B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable, subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate solution 24-hc'-rs is within the limits of Figure 3.1.7-1,.,
| |
| (continued)
| |
| Brunswick Unit 1 3.1-20 Amendment No. 2O2,
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate solution is 24-hetrs within the limits of Figure 3.1.7-2.
| |
| * SR 3.1.7.3 Verify temperature of pump suction and discharge/ 2A-heur-s SR 3.1.7.4 Verify continuity of explosive charge.
| |
| * SR 3.1.7.5 Verify the concentration of boron in so-uio1-witysL the limits of Figure 3.1.7-1.*,E' AND Once within In wth accodance the24 hours afterwae Contrl Prgramsolution temperature is restored within the limits of Figure 3.1.7-2 (continued)
| |
| Brunswick Unit 1 3.1-21 Bruswck ni I .121Amendment No. 22-7
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate Ž_41.2 gpm at a In accordance with discharge pressure > 1190 psig. the Inservice Testing Program SR 3.1.7.7 Verify flow through one SLC subsystem from pump 24-mont,,s-on
| |
| * into reactor pressure vessel. STAGGERED SR 3.1.7.8 Verify sodium pentaborate enrichment is > 47 ao Prior to addition to percent B-10. SLC tank I
| |
| IIn accordance with the Surveillance Frequency Control Program r
| |
| I Brunswick Unit 1 3.1-22 Bruswck ni I .122Amendment No. 227
| |
| | |
| SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1---------------------NOTE---------------
| |
| Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.
| |
| Verify each SDV vent and drain valve is open. 2A4-eays SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully 3!-days closed and fully open position.//
| |
| SR 3.1.8.3 Verify each SDV vent and drain valve: 2",cth
| |
| : a. Closes in < 30 seconds after receipt of a ra *ua or simulated scram signal; and V
| |
| : b. Opens when the actual or simulated s r -*
| |
| signal is reset. /
| |
| f In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.1-26 Brunsick Uit I .1-26Amendment No. O
| |
| | |
| APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER _>23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not A.1 Restore APLHGR(s) to 4 hours within limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS S URVEI LLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND 21 hou-r thereafter IIn accordance with the Surveillance Frequency Control Program jI II II Brunswick Unit 1 3.2-1 BrunwickUni I 32-1Amendment No. 2 MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| LCO 3.2.2 specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER _>23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A,.1 Restore MCPR(s) to within 4 hours limits, limits.
| |
| B. Required Action and 53.1 Reduce THERMAL 4 hours asoitdCompletion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified in the COLR. *12 hours after
| |
| Ž23% RTP AND 212, hourc thercaftcr (continued)
| |
| Brunswick Unit 1 3.2-2 Amendment No. 2-2 I
| |
| | |
| LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER >Ž23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž 23% RTP AND In accordance with the ~I Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.2-4 Bruswik Uit 3.-4Amendment No. 24
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES---------------
| |
| : 1. Refer to Table 3.3.1.1-1Ito determine which SRs apply for each RPS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.1 (Not used.)
| |
| SR 3.3.1.1.2 Perform CHANNEL CHECK. 24-heurs SR 3.3.1.1.3----------NOTE----------
| |
| Not required to be performed until 12 hours after THF:RMAI PO)WFR > 2)3o~/, IPTP Adjust the average power range monitor (APRM) channels to conform to the calculated power while operating at >_23% RTP.
| |
| SR 3.3.1.1.4 ------------- NOTE-------
| |
| Not required to be performed when entering MODI from MODE 1 until 12 hours after entering MODEi Perform CHANNEL FUNCTIONAL TEST.
| |
| '1 (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-4 BrunwickUni I 33-4Amendment No. 2 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform a functional test of each automatic scram 7dy contactor.
| |
| SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to withdrawing intermediate range monitor (IRM) channels overlap. SRMs from the fully Only bemet rquire dringsetryeitopMOD2tfro to MODE 1.
| |
| Verify the IRM and APRM channels overl* . -ay SR 3.3.1.1.8 Calibrate the local power range monit s.200cetifl I
| |
| SR 3.3.1.1.9 Perform CHANNEL FUNCTIONA TEST. 92dy SR 3.3.1.1.10 Calibrate the trip units. /_/_2-days...
| |
| I (continued)
| |
| In accordance with the Surveillance Frequency Control Program III I1*
| |
| Brunswick Unit 1 3.3-5 Bruswik Uit 3.-5Amendment No. 2-84
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.11 -~~~~NOTES----- --
| |
| : 1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
| |
| : 2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.
| |
| Perform CHANNEL FUNCTIONAL TEST. A 4-84-deys SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. S24 months SR 3.3.1.1.13 1. -~~~NOTES---
| |
| Neutron detectors are excluded.'- --
| |
| : 2. For Function 1, not required to be perf when entering MODE 2 from MODE 1 12 hours after entering MODE 2.
| |
| : 3. For Functions 2.b and 2.f, the recir transmitters that feed the APRMs Perform CHANNEL CALIBRATION. S2"!,months SR 3.3.1.1.14 (Not used.)
| |
| SR 3.3.1.1.15 Perform LOGIC SYSTEM FUN(
| |
| 2,A- months (continued)
| |
| In accordanceFrequency with the Surveillance Control Program l
| |
| I l
| |
| Brunswick Unit 1 3.3-6 Brunsick Uit 1 .3-6Amendment No. 24-I-
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve--Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure--Low Functions are not bypassed when THERMAL POWER '4 is >_26% RTP.
| |
| I SR 3.3.1.1.17 -~~~~NOTES-------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. For Functions 3 and 4, the sensor response timqi may be assumed to be the design sensor !
| |
| response time./
| |
| For Function 5, "n" equals I channels for the BASIS Frcquency.
| |
| For Function 2.e, "n" equals 8 channels for h D A C'IO I-'*
| |
| r'3*r.] If *+;*n I:
| |
| outputI., 4. sh,,.ll a'lte'rnate.
| |
| =
| |
| Verify the RPS RESPONSE TIME is within Ii 21! months on a STAGGERED
| |
| =
| |
| SR 3.3.1.1.18 Adjust recirculation drive flow to conform, core flow./ afer reaching equilibrium
| |
| ~'hce within 7 days conditions following refueling outage (continued)
| |
| In accordance with the Surveillance Frequency Control Program d
| |
| Brunswick Unit 1 3.3-7 Bruswik Uit 3.-7Amendment No. 2 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM Simulated 2A-,mciths Thermal Power is _>25% and recirculation drive flow is.,
| |
| < 60%. z In accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.3-8 BrunwickUni I 33-8Amendment No. 24-7
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS
| |
| -------- NOTE------------------------------...............
| |
| Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK.!2h-r SR 3.3.1.2.2----------NOTES---------
| |
| : 1. Only required to be met during CORE ALTERATIONS.
| |
| : 2. One SRM may be used to satisfy more than on*
| |
| of the following.
| |
| Verify an OPERABLE SRM detector is located in: 424hours
| |
| : a. The fueled region;,,
| |
| : b. The core quadrant where CORE ALTERATI N are being performed, when the associated R is included in the fueled region; and II (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-14 BrunwickUnitI No. 24-7 I 3.-14Amendment
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.4 ........................NOTES--------
| |
| : 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
| |
| : 2. Not required to be met during a core spiral offload.
| |
| Verify count rate is __3.0 cps.
| |
| ALTERATIONS AN4P 24-hc'-rs SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.1.2.6 ----------- NOTE Not required to be performed until 12 hoursj on Range 2 or below./
| |
| Perform CHANNEL FUNCTIONAL TES ! deys (continued)
| |
| L Surveillance Frequency In accordance Control with the Program Brunswick Unit 1 3.3-t5 Bruswik Uit No. 24-7 I 3.-15Amendment
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.7----------NOTES---------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
| |
| Perform CHANNEL CALIBRATION. 24-months I
| |
| Surveillance Frequency Control Program Brunswick Unit 1 3.3-16 No. 24 I BrunwickUnitI 3.-16Amendment
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor Mode E.1 Suspend control rod Immediately Switch--Shutdown Position withdrawal.
| |
| channels inoperable.
| |
| AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing onie or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
| |
| : 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
| |
| SURVEILLANCE/FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. .-1-84-1ay-s I (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 1 3.3-20 BruswikUit 3.-20Amendment No. 24-
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.2-----------NOTE- -------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn at < 8.75% RTP in MODE 2.
| |
| Perform CHANNEL FUNCTIONAL TEST. *'2-4a...
| |
| SR 3.3.2.1.3-------------------NOTE----------------
| |
| Not required to be performed until 1 hour after THERMAL POWER is *<8.75% RTP in MODE 1.
| |
| Perform CHANNEL FUNCTIONAL TEST. a 4..
| |
| SR 3.3.2.1.4 Verify the RBM: 2-~~h
| |
| : a. Low Power Range--Upscale Function 0 Intermediate Power Range--Upscale F ci OR High Power Range--Upscale Func o enabled (not bypassed) when APRM S ae Thermal Power is >_29%.
| |
| : b. Intermediate Power Range--Upscal Fnci OR High Power Range--Upscale F ini enabled (not bypassed) when APRI' mle Thermal Power is > Intermediate PF Rg Setpoint specified in the COLR.
| |
| : c. High Power Range--Upscale Fu i i enabled (not bypassed) when A imulated Thermal Power is _>High Poe e Setpoint specified in the COLR.
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-21 BrunwickUnitI 3.-2 1Amendment No. 22-2
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when THERMAL 24 mcnt~he POWER is
| |
| * 8.75% RTP.
| |
| SR 3.3.2.1.6-- - - - -- - - - -NOTE-.........................
| |
| Not required to be performed until 1 hour after reactorj mode switch is in the shutdown position. /
| |
| Perform CHANNEL FUNCTIONAL TEST. 24 mRths SR 3.3.2.1.7 -------- NOTE------------......
| |
| Neutron detectors are excluded.
| |
| Perform CHANNEL CALIBRATION. 2A-Renthe SR 3.3.2.1.8 Verify control rod sequences input to the conformance with BPWS. RWM OPERABLE folowing loading of
| |
| *ePIior to declaring squence into RM SIn accordance with the Surveillance Frequency Control Program II l
| |
| II Brunswick Unit 1 3.3-22 Bruswik Uit 3.-22Amendment No. 22-2
| |
| | |
| Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE REQUIREMENTS
| |
| ---------------- I'J I--
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided feedwater and main turbine high water level trip capability is maintained.
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL. CHECK. 2A-hour s aJ SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The Allowable A-months Value shall be < 207 inches.
| |
| SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, including valve actuation.
| |
| SIn accordanceFrequency Surveillance with the Control Program l
| |
| I m
| |
| Brunswick Unit 1 3.3-25 BrunwickUnitI 3.-25Amendment No. 2Oa
| |
| | |
| PAM Instrumentation 3.3.3.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| : 0. Required Action and 0.1 Enter the Condition Immediately associated Completion Time referenced in of Condition C not met. Table 3.3.3.1-1 for the channel.
| |
| E. As required by Required E.1 Be in MODE 3. 12 hours Action D.1 and referenced in Table 3.3.3.1-1.
| |
| F. As required by Required F.1 Initiate action in accordance Immediately Action D.1 and referenced in with Specification 5.6.6.
| |
| Table 3.3.3.1-1.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----------------
| |
| These SRs apply to each Function in Table 3.3.3.1-1.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK.3--dy Brunswick Unit 1 3.3-27 Amendment No. 2-34
| |
| | |
| PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.3 Perform CHANNEL CALIBRATION for each required 24-m,,c~ths PAM Instrumentation channel *_, I In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-28 Bruswck ni 1 .328Amendment No. 2-34
| |
| | |
| Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Remote Shutdown Monitoring Instrumentation LCO 3.3.3.2 The Remote Shutdown Monitoring Instrumentation Functions shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS
| |
| ------------- NC)
| |
| Separate Condition entry is allowed for each Function. I CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable, to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| S URVEI LLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required ,3- ,d...
| |
| instrumentation channel that is normally energized. ,,
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| /
| |
| Brunswick Unit 1 3.3-30 Bruswik Uit 3.-30Amendment No. 2-3-3
| |
| | |
| Remote Shutdown Monitoring Instrumentation 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each required 24 m,,cths instrumentation channel. =
| |
| SIn accordance with the Surveillance Frequency Control Program II l Brunswick Unit 1 3.3-31 Brunsick Uit I .3-31Amendment No. O
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One Function with B.1 Restore ATWS-RPT trip 72 hours ATWS-RPT trip capability capability.
| |
| not maintained.
| |
| C. Both Functions with C.1 Restore ATWS-RPT trip 1 hour ATWS-RPT trip capability capability for one Function.
| |
| not maintained.
| |
| D. Required Action and D.1 Remove the associated 6 hours associated Completion Time recirculation pump(s) from not met. service.
| |
| O__R D.2 Be in MODE 2. 6 hours SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----- ----------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 24-heut-s (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-33 Brunsick Uit I .3-33Amendment No. 2~
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.4.1.3 SR 3.3.4.1.4 Calibrate the trip units.
| |
| Perform CHANNEL CALIBRATION. The Allowable Q249 Values shall be:
| |
| : a. Reactor Vessel Water Level--Low Level 2:
| |
| >_101 inches; and
| |
| : b. Reactor Vessel Pressure--High: *<1147 psit SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST including breaker actuation. 24 meqths In accordance with the Surveillance Frequency Control Program d
| |
| Brunswick Unit 1 3.3-34 BrunwickUnitI 3.-34Amendment No. 2-a
| |
| | |
| ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS
| |
| ...................................... NOTES- - - - - - - - - - - - - - -
| |
| : 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c; and (b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
| |
| Brunswick Unit 1 3.3-40 Bruswik Uit 3.-40Amendment No. 2-§2
| |
| | |
| RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.
| |
| Brunswick Unit 1 3.3-47 Brunsick Uit I .3-47Amendment No. O
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES---------------
| |
| : 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions 2.c, 2.d, 3.a, 3.b, 3.e, 3.f, 3.g, 3.h, 4.a, 4.b, 4.e, 4.f, 4.g, 4.h, 4.i, 4.k, 5.a, 5.b, 5.e, 5.f, and 6.a; and (b) for up to 6 hours for all other Functions provided the associated Function maintains isolation capability.
| |
| Brunswick Unit 1 3.3-52 BrunwickUnitI No. 2-g$
| |
| 3.-52Amendment
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SU RVEI LLANCE IFREQ UENCY SR 3.3.6.1.8 ------------ NOTES---------
| |
| : 1. Radiation detectors are excluded.
| |
| : 2. The sensor response time for Functions l .a and 1.c may be assumed to be the design sensor response time.
| |
| Verify the ISOLATION INSTRUMENTATION RESPONSE TIME is within limits. ESTG C,,EIE SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST.
| |
| I In accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.3-53 Brunsick Uit I .3-53Amendment No. Q
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS ____________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 hour secondary containment isolation dampers inoperable.
| |
| AND C.2.1 Place the associated 1 hour standby gas treatment (SGT) subsystem(s) in operation.
| |
| O__R C.2.2 Declare associated SGT 1 hour subsystem(s) inoperable.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 24. hers (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program yI Brunswick Unit 1 3.3-60 Brunsick Uit I .3-60Amendment No. O
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92-dtays SR 3.3.6.2.3 Calibrate the trip unit. 2-In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-61 Brunswck Unt I 33-6 1Amendment No. O
| |
| | |
| CREV System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREV initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 24-.he'-rs SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. Q2dy SR 3.3.7.1.3 Perform CHANNEL CALIBRATION. , .A-.morths Brunswick Unit 1 3.3-64 Bruswck ni 1 .364Amendment No. 2<3
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Isolate condenser vacuum 12 hours associated Completion Time pumps.
| |
| of Condition A not met.
| |
| O__R OR B.2 Isolate main steam lines. 12 hours Condenser vacuum pump isolation capability not O.RR maintained.
| |
| B.3 Be in MODE 3. 12 hours SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation capability.
| |
| SU RVEI LLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK. a4-h9ur-s SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST.
| |
| ,Q2-onth SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The A Value shall be < 6 x background. /#
| |
| (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program I
| |
| Brunswick Unit 1 3.3-67 Bruswck ni I .367Amendment No. 2-*
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 2A-memthe including condenser vacuum pump trip breaker and isolation valve actuation.
| |
| IIn accordance with the Surveillance Frequency Control Program II I Brunswick Unit 1 3.3-68 Bruswck ni I .368Amendment No. 2-3
| |
| | |
| LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS
| |
| ..............--....................... NOTES- - - - - - - - - - - - - - -
| |
| : 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Functions maintains initiation capability for three DGs; and (b) for Function 2, the associated Function maintains DG initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST. 3-1-,4...
| |
| SR erformCHANNE.CALIBATION.48-meth 3.38.1.2 SR 3.3.8.1.3 Perform CHANNEL CALIBRATION. 18 -inths SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TES . 2A-mcFths IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-70 Bruswik Uit 3.-70Amendment No. 23Q
| |
| | |
| RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to fully insert Immediately associated Completion Time all insertable control rods in of Condition A or B not met core cells containing one or in MODE 3, 4, or 5 with any more fuel assemblies.
| |
| control rod withdrawn from a core cell containing one or more fuel assemblies.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.1 ------------ NOTE- -------
| |
| Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 for
| |
| _>24 hours.
| |
| Perform CHANNEL FUNCTIONAL TEST. 4~4-d~s SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS 2A. meRh:
| |
| motor generator set electric power monitoring assembly. The Allowable Values shall be:
| |
| : a. Overvoltage < 129 V.
| |
| : b. Undervoltage >_105 V.
| |
| : c. Underfrequency _>57.2 Hz.
| |
| (continued) lIn accordance with the II Surveillance Frequency Control Program If I
| |
| Brunswick Unit 1 3.3-73 BrunwickUnit1 3.-73Amendment No.* 2 I
| |
| | |
| RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS 2-4months alternate power supply electric power monitoring assembly. The Allowable Values shall be:
| |
| i,
| |
| : a. Overvoltage _<132 V.
| |
| : b. Undervoltage > 108 V.
| |
| : c. Underfrequency >_57.2 Hz.
| |
| SR 3.3.8.2.4 Perform a system functional test. I24-ment4h I
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.3-74 Bruswck ni 1 .374Amendment No. 2398
| |
| | |
| Recirculation Loops Operating 3.4.1 ACTIONS (continued) ________________
| |
| COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met.
| |
| O__R No recirculation loops in operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1------------NOTE---- -----
| |
| Not required to be performed until 24 hours after both recirculation loops are in operation.
| |
| Verify recirculation loop jet pump flow mismatch with 24-hetr-s both recirculation loops in operation:
| |
| : a. < 10% of rated core flow when operating at
| |
| < 75% of rated core flow; and
| |
| : b. < 5% of rated core flow when operating at /
| |
| _>75% of rated core flow.
| |
| In accordance with the Surveillance Frequency Control Program
| |
| * .-.---.-II Brunswick Unit 1 3.4-2 BrunwickUni I 34-2Amendment No. 2-44
| |
| | |
| Jet Pumps 3.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1-----------NOTES- -------
| |
| : 1. Not required to be performed until 4 hours after associated recirculation loop is in operation.
| |
| : 2. Not required to be performed until > 25% RTP.
| |
| Verify at least one of the following criteria (a or b) is 2-4-heu*s satisfied for each operating recirculation loop: *
| |
| : a. Recirculation pump flow to speed ratio differs by
| |
| <_5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by < 5% from established patterns.
| |
| : b. Each jet pump diffuser to lower plenum differential pressure differs by < 10% from that jet pump's established pattern.
| |
| In accordance with the l Surveillance Frequency Control Program Brunswick Unit 1 3.4-4 BrunwickUni I 34-4Amendment No. 24
| |
| | |
| SRVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.4.3.2-----------NOTE---.......................
| |
| Not required to be performed until 12 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify each required SRV opens when manually 2"-FPmonthS actuated.
| |
| IIn accordance with the Surveillance Frequency Control Program II l Brunswick Unit 1 3.4-6 Bruswik Uit 3.-6Amendment No. 2-2
| |
| | |
| RCS Operational LEAKAGE 3.4.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Pressure boundary LEAKAGE exists.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and 8-hei-s unidentified LEAKAGE increase are within limits. ,,
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.4-8 Brunsick Uit I .4-8Amendment No. O
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required primary 12-hours containment atmosphere radioactivity monitoring system.
| |
| SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of require*I leakage detection instrumentation. /,4 d~ys SR 3.4.5.3 .Perform a CHANNEL CALIBRATION of required leakage detection instrumentation.
| |
| Surveillance Frequency IIn accordance with the Control Program Ib I
| |
| Brunswick Unit 1 3.4-11 Bruswik Uit No. 2-g2 3.-11Amendment
| |
| | |
| RCS Specific Activity 3.4.6 ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours AND B.2.2.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1-------------NOTE--------
| |
| Only required to be performed in MODE 1.
| |
| Verify reactor coolant DOSE EQUIVALENT I-131 specific activity is < 0.2 IpCi/gm.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.4-13 Bruswik Uit 3.-13Amendment No. 2:Qg
| |
| | |
| RHR Shutdown Cooling System--Hot Shutdown 3.4.7 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.7.1-----------NOTE---------
| |
| Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR shutdown cooling isolation pressure.
| |
| Verify one required RHR shutdown cooling subsystem 42-heur-s or recirculation pump is operating.
| |
| In accordance with the Surveillance Frequency Control Program v/
| |
| q Brunswick Unit 1 3.4-16 Brunsick Uit 1 .4-16Amendment No. O
| |
| | |
| RHR Shutdown Cooling System--Cold Shutdown 3.4.8 ACTIONS (continued)_________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hour from subsystem in operation, circulating by an alternate discovery of no method. reactor coolant AND circulation No recirculation pump in AND operation.
| |
| Ornce per 12 hours thereafter AND B.2 Monitor reactor coolant Once per hour temperature.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling subsystem 4-2--heus or recirculation pump is operating.
| |
| SIn accordance with the II Surveillance Frequency Control Program p/
| |
| l Brunswick Unit 1 3.4-18 Brunsick Uit 1 .4-18Amendment No. Q
| |
| | |
| RCS P/T Limits 3.4.9 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C.------NOTE--...........C.1 Initiate action to restore Immediately Required Action 0.2 shall be parameter(s) to within completed if this Condition is limits.
| |
| entered.
| |
| -------- AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in'other than acceptable for operation. MODE 2 or 3.
| |
| MODES 1, 2, and 3.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.1---------------------NOTE---------------
| |
| Only required to be performed during RCS heatup and cooldown operations.
| |
| Verify: 3-mn,*if,, *,,
| |
| : a. RCS pressure and RCS temperature are within the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2; and
| |
| : b. RCS heatup and cooldown rates are < 100°F in any 1 hour period.
| |
| (continued)
| |
| Surveillance Frequency Control Program I In accordance with the Brunswick Unit 1 3.4-20 Brunsick Uit I .4-20Amendment No. 0
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY
| |
| +
| |
| SR 3.4.9.2 - -- - - - - NOTE.................
| |
| Only required to be performed during RCS inservice leak and hydrostatic testing.
| |
| Verify: 24-Riates
| |
| : a. RCS pressure and RCS temperature are within the applicable limits specified in Figure 3.4.9-3; 3.4.9-4, or 3.4.9-5, as applicable.
| |
| : b. RCS heatup and cooldown rates are < 30°F in any 1 hour period.
| |
| SR 3.4.9.3 Verify RCS pressure and RCS temperature are within nce within the criticality limits specified in Figure 3.4.9-2. i5 minutes prior to ov"ntrol rod ithdrawal for the purpose of achieving criticality SIR 3.4.9.4-----------NOTE-----------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the bottom head cool ntOnce within temperature and the reactor pressure vessel (RPV 30 minutes prior to coolant temperature is < 145°F. each startup of a S recirculation pump I]
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program m m Brunswick Unit 1 3.4-21 Uit 3.-21Amendment Bruswik No. 2-2-8
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS ('continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.5-----------NOTE----------------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the reactor coolant Once within temperature in~the recirculation loop to be started and 30 minutes prior, to the RPV coolant temperature is < 50°F. each startup of a recirculation pump SR 3.4.9.6-----------NOTE----------------
| |
| Only required to be performed when tensioning the reactor vessel head bolting studs.
| |
| Verify reactor vessel flange and head flange 8-iue temperatures are > 70°F.
| |
| SR 3.4.9.7-----------NOTE---------
| |
| Not required to be performed until 30 minutes after RCS temperature _<80°F in MODE 4.
| |
| Verify reactor vessel flange and head flange ~O-R~*Jtes temperatures are > 70°F.
| |
| I SR 3.4.9.8 ----------- NOTE--------
| |
| Not required to be performed until 12 hours after R(
| |
| temperature < 100°F in MODE 4.
| |
| Verify reactor vessel flange and head flange temperatures are _>70°F. 42-eJ
| |
| "" S*
| |
| Y In accordance with the Surveillance Frequency Control Program l
| |
| I dl Brunswick Unit 1 3.4-22 BrunwickUnitI3.-22Amendment No. 2-g8
| |
| | |
| Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be _<1045 psig.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is < 1045 psig. !2--he'-s I
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.4-28 BrunwickUnitI No. 22-8 I 3.-28Amendment
| |
| | |
| ECCS--Operating 3.5.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME J. Two or more low pressure J.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A or B.
| |
| OR HPCI System and two or more required ADS valves inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, the 34dy piping is filled with water from the pump discharge valve to the injection valve.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program t
| |
| II II Brunswick Unit 1 3.5-4 BrunwickUni I 35-4Amendment No. 2Oa-
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.2------------NOTE--------
| |
| Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling isolation pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each ECCS injection/spray subsystem manual, ~-1-,-d...
| |
| power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in /
| |
| position, is in the correct position.
| |
| SR 3.5.1.3 Verify ADS pneumatic supply header pressure is '*-1,-d...
| |
| > 95 psig.
| |
| SR 3.5.1.4 Verify the RHR System cross tie valve is locked 3 4..
| |
| closed.
| |
| SR 3.5.1.5------------NOTE--------
| |
| Not required to be performed if performed withi t previous 31 days.
| |
| Verify each recirculation pump discharge val a Once each startup bypass valve cycles through one complete leffl rior to exceeding travel or is de-energized in the closed posi 25% RTP (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program l
| |
| I II Brunswick Unit 1 3.5-5 Brunsick Uit I No. O
| |
| .5-5Amendment
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.6 Verify the following ECCS pumps develop the 92~de~,LS specified flow rate against a system head corresponding to the specified reactor pressure. #4 SYSTEM HEAD CORRESPONDI NC NO. OF TO A REACTOR.
| |
| SYSTEM FLOW RATE PUMPS PRESSURE OF I CS __4100gpm 1 ___113 psig LPCi > 14,000 gpm 2 >_20 psig SR 3.5.1.7 ------------ NOTE- -----
| |
| Not required to be performed until 48 hours aft*.
| |
| reactor steam pressure is adequate to perforrrl test. I Verify, withpump reactor Q2-daas the HPCI unitpressure < 1045 can develop and rat a flow Ž>*45 psig,
| |
| > 4250 gpm against a system head corre* onding tc reactor pressure./
| |
| SR 3.5.1.8 Not required to be performed until 48 ours reactor----------
| |
| steam pressure is NOTE adequate t / perfor test. I 2A-memths Verify, with reactor pressure < I1 pump unit can develop a flow rat a system head corresponding to (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.5-6 BrunwickUni I 35-6Amendment No. 24
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.9------------NOTE--------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each ECCS injection/spray subsystem actuates 2"!.meithe on an actual or simulated automatic initiation signal.
| |
| * SR3511-----------OTE-----------
| |
| Valve actuation may be excluded.
| |
| Verify the ADS actuates on an actual or simulated__2"A.menths automatic initiation signal. 1 SR 3.5.1.11-----------NOTE---------
| |
| Not required to be performed until 12 hours aft*
| |
| reactor steam pressure is adequate to perfor th test.
| |
| Verify each required ADS valve opens whe mauly2A-meih Verify the ECCS RE sPOS wTiME oreahCOS 24-meat-h injection/spray subsystemiswtn lmtz SSurveillance IIn accordanceFrequency with the Control Program Brunswick Unit 1 3.5-7 BrunwickUni I 35-7Amendment No.-2O*
| |
| | |
| ECCS-Shutdown 3.5.2 ACTIONS (continued) ________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 and D.1 Initiate action to restore Immediately associated Completion Time secondary containment to not met. OPERABLE status.
| |
| AND D.2 Initiate action to restore one Immediately standby gas treatment subsystem to OPERABLE status.
| |
| AND D.3 Initiate action to restore Immediately isolation capability in each required secondary containment penetration flow path not isolated.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure coolant injection !2-hc'-s (LPCI) subsystem, the suppression pool water level is In witacoranc th t*(continued)
| |
| Brunswick Unit 1 3.5-9 Amendment No. Q
| |
| | |
| ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 4
| |
| SR 3.5.2.2 Verify, for each required core spray (CS) subsystem, the:
| |
| : a. Suppression pool water level is >_-31 inches; or b.---------NOTE--- ---------- /
| |
| Only one required CS subsystem may take credit for this option during OPDRVs./
| |
| SR 3.5.2.3 Condensate storage tank water volume is
| |
| >_228,200 gallons.
| |
| Verify, for each required ECCS injection/spray I 2-ty subsystem, the piping is filled with water from the A pump discharge valve to the injection valve. !
| |
| SR 3.5.2.4------------NOTE----------------.....
| |
| One LPCI subsystem may be considered OPE BL during alignment and operation for decay heatrmoI if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each required ECCS injection/spray su s enb-~
| |
| manual, power operated, and automatic valv ith flow path, that is not locked, sealed, or otheri secured in position, is in the correct position (continued)
| |
| In accordance with the Surveillance Frequency Control Program
| |
| * J Brunswick Unit 1 3.5-10 BrunwickUnitI 3.-10Amendment No. 2-Q8
| |
| | |
| ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 92deL SR 3.5.2.5 Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure.
| |
| SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS > 113 psig
| |
| __4100gpm 1 LPCI > 9000 gpm I > 20 psig SR 3.5.2.6 ------------- NOTE--------------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each required ECCS injection/spray subsysten 14Rmonths actuates on an actual or simulated automatic initiatioi signal.
| |
| S R 3 .5 .2.7 - - - - - - - - - - - - - - - - - - - - - NOTE- - - - - - - - - -.--- - -
| |
| Instrumentation response time may be assumed to b/
| |
| the design instrumentation response time.
| |
| Verify the ECCS RESPONSE TIME for each require*
| |
| ECCS injection/spray subsystem is within the limit. 12A-meat-hs In accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 1 3.5-11 BrunwickUnitINo. 2§2, 3.-11Amendment
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water from -- de the pump discharge valve to the injection valve. 1 SR 3.5.3.2 Verify each RCIC System manual, power operated, 2.1-day and automatic valve in the flow path, that is not locke sealed, or otherwise secur~ed in position, is in the correct position.
| |
| SR 3.5.3.3------------NOTE--...................
| |
| : 1. Use of auxiliary steam for the performan o the SR is not allowed.
| |
| : 2. Not required to be performed until 24 o sate reactor steam pressure is adequate t prom the test.
| |
| Verify, with reactor pressure _Ž 945 psig d< 1045 -dS psig, the
| |
| Ž__400 gpmRCIC pump against can develop a system head acor*
| |
| fib ponditng ate to , "
| |
| reactor pressure. °7 (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.5-13 Bruswik Uit 3.-13Amendment No. 2-82
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.4 -~~~~NOTES--------
| |
| : 1. Use of auxiliary steam for the performance of the SR is not allowed with reactor pressure
| |
| > 150 psig.
| |
| : 2. Not required to be performed until 24 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify, with turbine inlet pressure Ž_135 psig and < 165 2A-Renthe psig, the RCIC pump can develop a flow rate
| |
| _Ž400 gpm against a system head corresponding to an.
| |
| equivalent reactor pressure.
| |
| I SR 3.5.3.5 -~~~~NOTE--------.
| |
| Vessel injection may be excluded. It Verify the RCIC System actuates on an actual or simulated automatic initiation signal.
| |
| IIn accordance with the Surveillance Frequency Control Program I
| |
| I II Brunswick Unit 1 3.5-14 Brunsick Uit I .5-14Amendment No. O
| |
| | |
| Primary Containment 3.6.1.1 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and leakage rate In accordance with testing, except for primary containment air lock testing, the Primary in accordance with the Primary Containment Leakage Containment Rate Testing Program. Leakage Rate Testing Program SR 3.6.1.1 .2 Verify drywell to suppression chamber differential 4-ment-hs pressure does not decrease at a rate > 0.25 inch water iA gauge per minute tested over a 10 minute period at an initial differential pressure ofŽ> 1.00 psid and
| |
| * 1.25 psid.*-
| |
| I n accordance with the Surveillance Frequency Control Program t/
| |
| Brunswick Unit 1 3.6-2 Brunsick Uit I .6-2Amendment No. -
| |
| | |
| Primary Containment Air Lock 3.6.1.2 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.2.1-----------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 3.6.1.1.1.
| |
| Perform required primary containment air lock leakage In accordance with rate testing in accordance with the Primary the Primary Containment Leakage Rate Testing Program. Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary containment air / mcnteRhs lock can be opened at a time. I Inaccordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.6-6 Amendment No. 2O2,
| |
| | |
| PClVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 -...............
| |
| NOTES---------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment isolation manual S1-days valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. /
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program
| |
| ¢'
| |
| Brunswick Unit 1 3.6-11 Brunsick Uit I .6-11Amendment No. O
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.2 ------------ NOTES- -------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing incore probe (TIP) av, 34....~
| |
| shear isolation valve explosive charge. I SR 3.6.1.3.4 Verify the isolation time of each power operated and Inraccordance each automatic PCIV, except for MSIVs, is within with the Inservice limits. Testing Program SR 3.6.1.3.5 Verify the isolation time of each MSIV is _> 3 secotd In accordance with and *<5 seconds. the Inservice Testing Program Inaccordance with the (cniud Surveillance Frequency Control Program Brunswick Unit 1 3.6-12 Amendment No. 24*
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 1ieth position on an actual or simulated isolation signal.
| |
| SR 3.6.1.3.7 Verify a representative sample of reactor I2A-R~eRthe instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal.
| |
| SR 3.6.1.3.8 Remove and test the explosive squib from each r'the n accordance isolation valve of the TIP System. Inservice with Testing Program SR 3.6.1.3.9 Verify leakage rate through each main ste*
| |
| < 100 scfh and the combined leakage rate the Primary main steam lines is _<150 scfh when testec. Containment SIn accordance with
| |
| > 25 psig. Leakage Rate Testing Program In accordance with the Surveillance Frequency Control Program 'i d
| |
| Brunswick Unit 1 3.6-13 Bruswck ni I .613Amendment No. 2-39
| |
| | |
| Drywell Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Air Temperature LCO 3.6.1.4 Drywell average air temperature shall be _<150°F.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell average air 8 hours temperature not within limit, temperature to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE/FREQUENCY SR 3.6.1.4.1 Verify drywell average air temperature is within limit. 24he*
| |
| IControl Program Brunswick Unit 1 3.6-14 Brunsick Uit I .6-14Amendment No. O
| |
| | |
| Reactor Building-to-Suppression Chamber Vaccum Breakers 3.6.1.5 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify nitrogen bottle supply pressure of each nitrogen 2A-hetrs backup subsystem is _>1130 psig. _
| |
| SR 3.6.1.5.2 ----------- NOTES--------
| |
| : 1. Not required to be met for vacuum breakerst are open during Surveillances./
| |
| : 2. Not required to be met for vacuum breaker/
| |
| open when performing their intended on Verify each vacuum breaker is closed.
| |
| SR 3.6.1.5.3 Perform a functional test of each vacuum_
| |
| SR 3.6.1.5.4 Verify the f 2A-R~eRth~
| |
| _<0.5 psid.
| |
| SR 3.6.1.5.5 Verify leakage rate of each nitroger subsystem is < 0.65 scfm when tes 24 e*. t4h e, nitrogen bottle supply pressure of j SR 3.6.1.5.6 Verify the Nitrogen Backup Sy I2A-meR~hs to the vacuum breakers on an actuation signal. j Surveillance Frequency In accordance with the Control Program I
| |
| Brunswick Unit 1 3.6-17 Brunsick Uit I .6-17Amendment No. 2~
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 -NOTE Not required to be met for vacuum breakers that are open during Surveillances.
| |
| Verify each vacuum breaker is closed.
| |
| Within 6 hours after any discharge of steam to the suppression chamber from any source AND Within 6 hours following an operation that causes any of the vacuum breakers to open SR 3.6.1.6.2 Perform a functional test of each rq §2-de~s breaker.
| |
| AND Within 12 hours after any SIn accordance with the discharge of Surveillance Frequency steam to the Control Program suppression II chamber from the SRVs SR 3.6.1.6.3 Verify the full open setpoint of each required vaum 2~4-imiths breaker is < 0.5 psid.
| |
| Brunswick Unit 1 3.6-19 BrunwickUnitI No. 2&-84 I 3.-19Amendment
| |
| | |
| Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature is within the applicable limits. A SAND 5 minutes when performing testing that adds heat to the suppression pool
| |
| /
| |
| In accordance with the Surveillance Frequency Control Program
| |
| /
| |
| Brunswick Unit 1 3.6-22 Brunsick Uit I .6-22Amendment No. O
| |
| | |
| Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be > -31 inches and _<-27 inches.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression pool 2 hours level not within limits, water level to within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits. 2-e*
| |
| SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.6-23 Brunsick Uit I .6-23Amendment No. O
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pooi cooling subsystem ~1d~
| |
| manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise /
| |
| while operating in the suppression pool cooling oe In accordance with the Surveillance Frequency Control Program iy I
| |
| Brunswick Unit 1 3.6-25 Brunsick Uit I .6-25Amendment No. O
| |
| | |
| Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LCO 3.6.3.1 The primary containment oxygen concentration shall be < 4.0 volume percent.
| |
| APPLICABILITY: MODE 1 during the time period:
| |
| : a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to'
| |
| : b. 24 hours prior to a scheduled reduction of THERMAL POWER to
| |
| < 15% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment oxygen A.1 Restore oxygen 24 hours concentration not within concentration to within limit.
| |
| limit.
| |
| B. Required Action and B.1 Reduce THERMAL 8 hours associated Completion Time POWER to _<15% RTP.
| |
| not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.'1.1 Verify primary containment oxygen concentration is 7-41ays within limits.
| |
| Surveillance Frequency Control Program Brunswick Unit 1 3.6-26 Amendment No. 243
| |
| | |
| Secondary Containment 3.6.4.1 ACTIONR£ COMPLETION CONDITION REQUIRED ACTION TIME C. (continued) C.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment equipment hatches 2A.-iciths are closed and sealed.
| |
| SR 3.6.4.1.2 Verify one secondary containment access door is ~A-meRths closed in each access opening.
| |
| SR 3.6.4.1.3 Verify each SGT subsystem can maintain > 0.25 incl S/rT'4,AGG ERED of vacuum water gauge in the secondary cnane for 1 hour at a flow rate<* 3000 cfm. I I
| |
| Iin accordance with the Surveillance Frequency Control Program l
| |
| * I Brunswick Unit 1 3.6-29 Bruswik Uit 3.-29Amendment No. 252 I
| |
| | |
| SCl~s 3.6.4.2 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME D. Required Action and D.1 -NOTE----
| |
| associated Completion Time LCO 3.0.3 is not applicable.
| |
| of Condition A or B not met - - - -
| |
| during movement of recently irradiated fuel assemblies in Suspend movement of Immediately the secondary containment recently irradiated fuel or during OPDRVs. assemblies in the secondary containment.
| |
| AND D.2 Initiate action to suspend Immediately OPDRVs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.2.1 Verify the isolation time of each automatic SCID is 2-eth within limits.
| |
| lI A
| |
| SR 3.6.4.2.2 Verify each automatic SCID actuates to the isolation position on an actual or simulated actuation signal.
| |
| In accordance with the Surveillance Frequency Control Program I
| |
| J Brunswick Unit 1 3.6-32 Bruswik Uit 3.-32Amendment No. 2-* I
| |
| | |
| SGT System 3.6.4.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two SGT subsystems E.1I-NOTE----.....
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of recently irradiated fuel - - -
| |
| assemblies in the secondary containment or during Suspend movement of Immediately OPDRVs. recently irradiated fuel assemblies in secondary containment.
| |
| AND E.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for _ 10 continuous "3 -d ...
| |
| hours with heaters operating.
| |
| SR 3.6.4.3.2 Perform required SGT filter testing in accordance In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.6.4.3.3 2A-m......hs Verify eachinitiation simulated signal. actuates on an actual c SGT subsystem IIn accordance with the Surveillance Frequency Control Program I
| |
| I I
| |
| Brunswick Unit 1 3.6-35 Bruswik Uit 3.-35Amendment No. 252-
| |
| | |
| RHRSW System 3.7.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION !COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual, power operated, and automatic valve in the flow path, that is not locked, I, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.
| |
| SIn accordance with the Surveillance Frequency Control Program U
| |
| Brunswick Unit 1 3.7-3 BrunwickUni I 37-3Amendment No. 2Oa-
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the SW pump suction bay of 2A-.hc,-r the intake structure is > -6 ft mean sea level. ,
| |
| SR 3.7.2.2 Verify the water temperature of UHS is _<90.5°F.
| |
| t-SR 3.7.2.3 -------- NOTE---------------........
| |
| Isolation of flow to individual components does not j render SW System inoperable. I Verify each SW System manual, power opera -4e~s automatic valve in the flow paths servicing sa related systems or components, that is not Io*
| |
| sealed, or otherwise secured in position, is in correct position.
| |
| SR 3.7.2.4 ----------- NOTES------
| |
| : 1. A single test at the specified Frequenc*
| |
| satisfy this Surveillance for both units.A
| |
| : 2. Isolation of flow to individual compor not render SW System inoperable.
| |
| Verify automatic transfer of each DG coo Q2-deaLs supply from the normal SW supply to th supply on low OG jacket cooling water pressure.I (continued)
| |
| In accordance with the Surveillance Frequency Control Program d
| |
| Brunswick Unit 1 3.7-9 Bruswik Uit 3.-9Amendment N o.-2-1-,
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.5-----------NOTE---------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each required SW System automatic component 2!.4-menth*
| |
| actuates on an actual or simulated initiation signal. ,,,
| |
| In accordance with the Surveillance Frequency Control Program p/
| |
| Brunswick Unit 1 3.7-10 BrunwickUnitI 3.-10Amendment No. 203
| |
| | |
| CREV System 3.7.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two CREV subsystems--------NOTE-------
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of irradiated fuel assemblies- ------------
| |
| in the secondary E.1 Suspend movement of Immediately containment, during CORE irradiated fuel assemblies ALTERATIONS, or during in the secondary OPDRVs. containment.
| |
| OR AND On rmrRVE.2 Suspend CORE Immediately subsystems inoperable due ATRTOS to an inoperable CRE boundary during movement AND of irradiated fuel assemblies E.3 Initiate action to suspend Immediately in the secondaryOPRs containment, during COREOPRs ALTERATIONS, or during OPD RVs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREV subsystem for _ 15 continuous 21d*
| |
| minutes.
| |
| SR 3..3.2 Perform required CREV filter testing in accordance* In accordance with SR.7..2 with the Ventilation Filter Testing Program (VFTP)./ the VFTP (continued)
| |
| Su nt re*l pnCg rFreq u e n cY I--t In accordance with the I /
| |
| Brunswick Unit 1 3.7-13 BruswikUit 3.-13Amendment No. 248
| |
| | |
| CREV System 3.7.3 SURVEILLANCE REQUI REMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.3 Perform required CRE unfiltered air inleakage testing In accordance with in accordance with the Control Room Envelope the Control Room Habitability Program. Envelope Habitability Program SR 3.7.3.4 Verify each CREV subsystem actuates on an actual or 24-moinths simulated initiation signal. .
| |
| In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.7-14 Bruswik Uit 3.-14Amendment No. 248
| |
| | |
| Control Room AC System 3.7.4 ACTIONS (continued)_________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME F. Three control room AC---------------NOTE---------
| |
| subsystems inoperable LCO 3.0.3 is not applicable.
| |
| during movement of irradiated fuel assemblies in the secondary containment, F.1 Suspend movement of Immediately during CORE irradiated fuel assemblies in ALTERATIONS, or during the secondary containment.
| |
| OP DRVs.
| |
| AND F.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND F.3 Initiate actions to suspend Immediately OPDR Vs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.'7.4.1 Verify each control room AC subsystem has the2",m th capability to remove the assumed heat load.
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.7-17 BrunwickUnitINo. 24O3 3.-17Amendment
| |
| | |
| Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4
| |
| SR 3.7.5.1 -~~~~~NOTE---------
| |
| Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
| |
| Verify the gross gamma activity rate of the noble 31 dacys gases is *<243,600 pCi/second otter decay ot AND 30 minutes.
| |
| Once within 4 hours after aŽ>50%
| |
| increase in the nominal steady state fission gas In accordance with the release after
| |
| ,Surveillance Frequency factoring out Control Program increases due to l changes in THERMAL POWER level Brunswick Unit 1 3.7-19 Bruswik Uit No. 2-O2 3.-19Amendment
| |
| | |
| Main Turbine Bypass System 3.7.6 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete cycle of each main turbine bypass 34-ey valve.
| |
| SR 3.7.6.2 Perform a system functional test. 2-iFh SR 3.7:6.3 Verify the TURBINE BYPASS SYSTEM RESPON TIME is within limits.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.7-21 BrunwickUnitI 3.-2 1Amendment No. 24*
| |
| | |
| Spent Fuel Storage Pool Water Level 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Spent Fuel Storage Pool Water Level LCO 3.7.7 The spent fuel storage pool water level shall be _ 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage racks.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.
| |
| ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool A.1------------NOTE------
| |
| water level not within limit. LCO 3.0.3 is not applicable.
| |
| Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the spent fuel storage pool water level is
| |
| _ 19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage racks. /.
| |
| Brunswick Unit 1 3.7-22 Amendment No. O
| |
| | |
| AC Sources--Operating 3.8.1 SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each offsite circuit.
| |
| '4 SR 3.8.1.2
| |
| : 1. -------
| |
| All DG starts mayNOTES---------
| |
| prelube period.
| |
| be preceded by an engine
| |
| //
| |
| : 2. A modified DG start involving idling and gra, acceleration to synchronous speed may be for this SR. When modified start procedure:.
| |
| not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
| |
| : 3. A single test at the specified Frequency41 satisfy this Surveillance for both units.
| |
| Verify achieveseachsteady DG starts state from 4 standby conditio and voltage >Ž3750 V and_*4300 V and frequency _>58.8 Hz and <*61.2 Hz.
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.8-7 Bruswik Uit 3.-7Amendment No. 26
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.3 S~~~~NOTES--------
| |
| : 1. DG loadings may include gradual loading.
| |
| : 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.
| |
| : 5. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG is synchronized and loaded and 31-deys operates for _ 60 minutes at a load > 2800 kW and
| |
| _<3500 kW. "
| |
| SR 3.8.1.4 Verify each engine mounted tank contains >_150 fuel oil. I aldy SR 3.8.1.5 Check engine for and remove mounted tank. accumulated water fr SR 3.8.1.6 Verify the fuel oil transfer system orc fuel oil from the day fuel oil storage mounted tank.
| |
| (continued)
| |
| Surveillance Frequency SIn accordance Control Program with the Ib I
| |
| 4 Brunswick Unit 1 3.8-8 Bruswik No. 2-0 I Uit 3.-8Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.7 ------------ NOTES---------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| IA Verify each OG starts from standby condition and achieves, in < 10 seconds, voltage Ž_3750 V and frequency_> 58.8 Hz, and after steady state conditions are reached, maintains voltage _>3750 V and
| |
| * 4300 V and frequency > 58.8 Hz and _<61.2 Hz.
| |
| -. 5.
| |
| (continued) lIn accordance with the Surveillance Frequency Control Program
| |
| * i Brunswick Unit 1 3.8-9 BrunwickUni I No. 2-06 I 38-9Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.8 ----------- NOTES- -------
| |
| : 1. SR 3.8.1 .8.a shall not be performed in MODE 1 or 2 for the Unit 1 offsite circuits. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. SR 3.8.1.8.a is not required to be met if the unit power supply is from the preferred offsite circuit.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify:
| |
| : a. Automatic transfer capability of the unit power supply from the normal circuit to the preferred 2AReV h offsite circuit; and
| |
| : b. Manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsitei circuit. /
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program III Brunswick Unit 1 3.8-10 BrunwickUnitI 3.-10Amendment No. 2-g8
| |
| | |
| AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.9 -~~~NOTES--------
| |
| : 1. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. If performed with the DG. synchronized with offsite power, it shall be performed at a power factor < 0.9.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG rejects a load greater than or equal to its associated core spray pump without tripping. 2A- eth (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program 1 1 Brunswick Unit 1 3.8-11 Bruswik Uit 3.-11Amendment No. 2-
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.10 -~NOTE---------------
| |
| A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG's automatic trips are bypassed on an
| |
| /
| |
| actual or simulated ECCS initiation signal except:
| |
| : a. Engine overspeed;
| |
| : b. Generator differential overcurrent;
| |
| : c. Low lube oil pressure; 1
| |
| : d. Reverse power;
| |
| : e. Loss of field; and
| |
| : f. Phase overcurrent (voltage restrained).
| |
| (continued)
| |
| Control Program Brunswick Unit 1 3.8-12 BrunwickUnitI No. 2-88 I 3.-12Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.11 ----------- NOTES- -------
| |
| : 1. Momentary transients outside the load and power factor ranges do not invalidate this test.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG operating at a power factor _<0.9 operates for > 60 minutes loaded to >_3500 kW and
| |
| < 3850 kW. 4 A4m~4 SR 3.8.1.12 A single test at the specified Frequency will satisfy thi
| |
| -------- NOTE---------
| |
| Surveillance for both units.]
| |
| Verify an actual or simulated ECCS initiation signal~ls
| |
| ~4-R~eHthe capable of overriding the test mode feature to retuf each DG to ready-to-load operation.
| |
| (continued) i IIn accordanceFrequency Surveillance with the Control Program l
| |
| Brunswick Unit 1 3,8-13 Bruswik Uit 3.-13Amendment No.
| |
| * AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.13--------------------NOTE----------------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 1 and OG 2. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify interval between each sequenced load block is 2A-,months within + 10% of design interval for each load sequence IJ relay.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program
| |
| * I Brunswick Unit 1 3.8-14 Bruswik Uit 3.-14Amendment No. 2- I
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.14-----------NOTES- -------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 1 and DG 2. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify, on actual or simulated loss of offsite power 2".mth signal initiationin conjunction signal: with an actual or simulated ECOS
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses; and
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in
| |
| < 10.5 seconds,
| |
| : 2. energizes auto-connected emergency loads through load sequence relays,
| |
| : 3. maintains steady state voltage _>3750 V and <*4300 V,
| |
| : 4. maintains steady state frequency > 58.8 Hz and
| |
| * 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected emergency loads for
| |
| __5 minutes.
| |
| SIn accordance with the I Surveillance Frequency Control Program
| |
| * II Brunswick Unit 1 3.8-15 BrunwickUnitI 3.-15Amendment No. 2 6
| |
| | |
| Diesel Fuel Oil 3.8.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.3.1 For each required DG, verify:v, .. *
| |
| : a. The associated day fuel oil storage tank contains Ž_22,650 gal; and
| |
| : b. The main fuel oil storage tank contains >_20,850 gal per required DG.
| |
| SR 3.8.3.2 Verify fuel oil properties of stored fuel oil are tested ii In accordance with accordance with, and maintained within the limits of the Diesel Fuel Oil the Diesel Fuel 0il Testing Program. / STesting Program I
| |
| SR 3.8.3.3 Check for and remove accumulated water from e*
| |
| day fuel oil tank and the main fuel oil storage tang m m SIn accordanceFrequency with the Surveillance Control Program
| |
| =
| |
| Brunswick Unit 1 3.8-22 BrunwickUnitI No. 2O 3.-22Amendment I
| |
| | |
| DC Sources--Operating 3.8.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Two or more DC electrical power subsystems inoperable.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is _>130 V on float 7-4a~&
| |
| charge.
| |
| /
| |
| SR 3.8.4.2 Verify no visible corrosion at battery terminals and connectors.
| |
| O__R Verify battery connection and resistance is _<23.0 p ohn
| |
| _< 82.8 tpohms for int1 for inter-cell connections rack connections.I SR 3.8.4.3 Verify battery cells, cell plates, and racks show !8-Imenths visual indication of physical damage or abnormj deterioration that degrades performance. /
| |
| (continued)
| |
| SIn accordanceFrequency Surveillance with the Control Program fl I
| |
| I Brunswick Unit 1 3.8-24 BrunwickUnit1 No. 2O I 3.-24Amendment
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell !8-imcnths and terminal connections are coated with anti-corrosion material./
| |
| SR 3.8.4.5 Verify each required battery charger supplies /24-menths
| |
| Ž>250 amps at ___135 V for Ž>4 hours.im, SR 3.8.4.6 ----------- NOTES--------
| |
| : 1. The modified performance discharge test in SR 3.8.4.7 may be performed in lieu of the service test in SR 3.8.4.6 once per 60 months.
| |
| : 2. This Surveillance shall not be performed in MULJL 1 or 2 tor 1 UUL electrical pov thle unitcredit subsystems. However, may be taken unplanned events that satisfy this SR.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify battery capacity is adequate to supply, and I24 mcn~he maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.8-25 Brunsick Uit I .8-25Amendment No.20
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.7 -~~~~NOTES---------
| |
| : 1. This Surveillance shall not be performed in MODE 1 or 2 for the Unit I DC electrical power subsystems. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify battery capacity is _Ž80% of the manufacturer's *0 Rmenths rating when subjected to a performance discharge tesp or a modified performance discharge test. /
| |
| 12 months when battery shows degradation or has reached 85% of the expected life with IIn accordance with the capacity < 100% of ISurveillance Frequency
| |
| * manufacturer's IControl Program rating A._ND 24 months when battery has reached 85% of the expected life with capacity Ž_100% of manufacturer's rating L ___________________________
| |
| Brunswick Unit 1 3.8-26 BrunwickUnitI No. 205 I 3.-26Amendment
| |
| | |
| Battery Cell Parameters 3.8.6 ACTIONS ____________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.
| |
| B. Required Action and B.1 Declare associated battery 'Immediately
| |
| .associated Completion Time inoperable.
| |
| of Condition A not met.
| |
| OR One or more batteries with average electrolyte temperature of the representative cells not within limits.
| |
| OR One or more batteries with one or more battery cell parameters not within Category C limits.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 7-d4ay Category A limits.
| |
| (continued)
| |
| LSurveillance Frequency Inaccordance Control Program with the j
| |
| Brunswick Unit 1 3.8-31 Brunsick Uit No. O I I .8-31Amendment
| |
| | |
| Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 "2-a" Category B limits.
| |
| SR 3.8.6.3 Verify average electrolyte temperature of representative cells is >_60°F.
| |
| In accordanceFrequency with the Surveillance Control Program I
| |
| l Brunswick Unit 1 3.8-32 BrunwickUnitI No. 29 I 3.-32Amendment
| |
| | |
| Distribution Systems--Operating 3.8.7 SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and indicated power 7dy availability to required AC and DC electrical power ,
| |
| distribution subsystems.
| |
| SR 3.8.7.2 Verify no combination of more than two power 7dy conversion modules (consisting of either two lighi inverters or one lighting inverter and one plant uninterruptible power supply unit) are aligned to Division II bus B.
| |
| In accordance with the Surveillance Frequency Control Program
| |
| _//
| |
| Brunswick Unit 1 3.8-37 BrunwickUnitI 3.-37Amendment No. 2O I
| |
| | |
| Distribution Systems-Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to suspend Immediately operations with a potential for draining the reactor vessel.
| |
| AND
| |
| *A.2.4 Initiate actions to restore Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and indicated power 7-d*
| |
| availability to required AC and DC electrical power distribution subsystems. 7 In accordance with the Surveillance Frequency Control Program
| |
| _/
| |
| Brunswick Unit 1 3.8-39 Bruswik Uit 3.-39Amendment No. 2O6
| |
| | |
| Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of 7-deys the following required refueling equipment interlock inputs:
| |
| : a. All-rods-in,
| |
| : b. Refuel platform position,
| |
| : c. Refuel platform fuel grapple, fuel loaded,
| |
| : d. Fuel grapple position,
| |
| /
| |
| : e. Refuel platform frame-mounted hoist, fuel loaded, and
| |
| : f. Refuel platform monorail hoist, fuel loaded.
| |
| In accordance with the Surveillance Frequency Control Program 1z/
| |
| ,I Brunswick Unit 1 3.9-2 BrunwickUni I 39-2Amendment No. 202,
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
| |
| ACTIONS_________________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-out A.1 Suspend control rod Immediately interlock inoperable, withdrawal.
| |
| AND A.2 Initiate action to fully insert Immediately all insertable control rods in
| |
| *core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in Refuel position. !2-liurs (continued) lIn accordance with the Surveillance Frequency yU Control Program Brunswick Unit 1 3.9-3 Brunsick Uit I .9-3Amendment No. O
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| S URVE ILLANCE FREQUENCY SR 3.9.2.2-------------NOTE---------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn.
| |
| Perform CHANNEL FUNCTIONAL TEST. ~ dy A
| |
| I SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.9-4 Brunsick Uit I No. O
| |
| .9-4Amendment
| |
| | |
| Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
| |
| APPLICABILITY: When loading fuel assemblies into the core.
| |
| ACTIONS_________________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control rods A.1 Suspend loading fuel Immediately not fully inserted, assemblies into the core.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE JFREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted. 42-hcurc In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.9-5 Brunsick Uit I .9-5Amendment No. O
| |
| | |
| Control Rod OPERABILITY--Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY--Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.
| |
| APPLICABILITY: MODE 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to fully insert Immediately control rods inoperable, inoperable withdrawn control rods.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1-------------NOTE----------------........
| |
| Not required to be performed until 7 days after the control rod is withdrawn.
| |
| Insert each withdrawn control rod at least one notch. ~-d4ays SR 3.9.5.2 Verify each withdrawn control rod scram accumulato/r pressure is > 940 psig. /f V IIn accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.9-8 Brunsick Uit I No. o
| |
| .9-8Amendment
| |
| | |
| RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 RPV water level shall be _> 23 ft above the top of irradiated fuel assemblies seated within the RPV.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies within the RPV, During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.
| |
| ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not within A.1 Suspend movement of fuel Immediately limit. assemblies and handling of control rods within the RPV.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is >_23 ft above the top of 2'!.he-rs irradiated fuel assemblies seated within the RPV. t I
| |
| In accordance with the Surveillance Frequency Control Program t_/
| |
| I Brunswick Unit 1 3.9-9 BrunsickUit I .9-9Amendment No.
| |
| | |
| RHR-High Water Level 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is 42-hc'-r,,
| |
| operating. j SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.9-12 Bruswik Uit 3.-12Amendment No. 2-
| |
| | |
| RHR--Low Water Level 3.9.8 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is !-2-hc,-r operating.
| |
| Control Program Brunswick Unit 1 3.9-15 Brunsick Uit I .9-15Amendment No. 2~
| |
| | |
| Reactor Mode Switch Interlock Testing 3.10.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Place the reactor mode 1 hour switch in the shutdown position.
| |
| OR A.3.2------NOTE----
| |
| Only applicable in MODE 5.
| |
| Place the reactor mode 1 hour switch in the refuel position.
| |
| SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted in core cells 42-he~f S containing one or more fuel assemblies.
| |
| SR 3.10.2.2 Verify no CORE ALTERATIONS are in 2A-.ho'-s I
| |
| In accordance with the Surveillance Frequency Control Program m
| |
| Brunswick Unit 1 3,10-5 BrunwickUnitI 3.0-5Amendment No. 2-Q8
| |
| | |
| Single Controi Rod Withdrawal--Hot Shutdown 3.10.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required LCOs. According to the Sapplicable SRs SR 3.10.3.2 ----------- NOTE- --------
| |
| Not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d. 1 requirements.
| |
| Verify all control rods, other than the control rod being 24-heur-s withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| SR 3.10.3.3 Verify all control rods, other than the control rod be 2A4-heu e withdrawn, are fully inserted.
| |
| lIn accordance with the Surveillance Frequency Control Program l
| |
| I II Brunswick Unit 1 3.10-8 Brunswck Unt I 310-8Amendment No. 2~
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 ACTIONS (continued)__________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One or more of the above B.1 Suspend withdrawal of the Immediately requirements not met with control rod and removal of the affected control rod not associated CRD.
| |
| insertable.
| |
| AND B.2.1 Initiate action to fully insert Immediately all control rods.
| |
| O__R B.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.4.2----------NOTE- --------
| |
| Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c. 1 requirements.
| |
| Verify all control rods, other than the control rod being 24-heI4F withdrawn, in a five by five array centered on the ,
| |
| control rod being withdrawn, are disarmed. 1
| |
| _/
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.10-11 BrunwickUnitI 3.0-11Amendment No. 2Q*
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS SR 3.10.4.3 Verify all control rods, other than the control rod being 24-heuws withdrawn, are fully inserted.
| |
| I. -
| |
| W ;
| |
| SR 3.10.4.4 Not required to be met if SR 3.10.4.1 is satisfiedfo/
| |
| LCO 3.10.4.b.1
| |
| -- -- -- OTE-----------Io requirements./
| |
| Verify a control rod withdrawal block is inserted. 24-heur-s In accordance with the Surveillance Frequency Control Program Brunswick Unit 1 3.10-12 Brunwic Unt I3.1-12Amendment No. 24§
| |
| | |
| Single CR0 Removal--Refueling 3.10.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Initiate action to fully insert Immediately all control rods.
| |
| OR A.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods, other than the control rod 24-heurs withdrawn for the removal of the associated CR0, arei1t fully inserted.
| |
| SR 3.10.5.2 Verify all control rods, other than the control rod 21-.hc'-rs withdrawn for the removal of the associated CR0, i ,a five by five array centered on the control rod withd awn for the removal of the associated CR0, are disarred SR 3.10.5.3 Verify a control rod withdrawal block is inserte 2A-heius SR 3.10.5.4 Perform SR 3.1.1.1..**kcrigt (continued)
| |
| In accordance with the Surveillance Frequency Control Program F-f,,
| |
| l Brunswick Unit 1 3.10-14 Brunwic Unt I3.1-14Amendment No. 282
| |
| | |
| Single CRD Removal--Refueling 3.10.5 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.5 Verify no other CORE ALTERATIONS are in progress. 2A-heur-s I
| |
| SIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 1 3.10-15 Brunwic Unt I3.1-15Amendment No. 2O2,
| |
| | |
| Multiple Control Rod Withdrawal--Refueling 3.10.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Initiate action to fully insert Immediately all control rods in core cells containing one or more fuel assemblies.
| |
| O__R A.3.2' Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.6.1 Verify the four fuel assemblies are removed from core 2!hus cells associated with each control rod or CRD removed.
| |
| SR 3.10.6.2 Verify all other control rods in core cells containing 01 2A, hours or more fuel assemblies are fully inserted.
| |
| SR 3.10.6.3 ----------- NOTE-------
| |
| Only required to be met during fuel loading. ,
| |
| Verify fuel assemblies being loaded are in ,2A-.he'ds with an approved spiral reload sequence.
| |
| IIn accordanceFrequency Surveillance with the Control Program l
| |
| II Brunswick Unit 1 3.10-17 Brunwic Unt I3.1-17Amendment No. 2-00
| |
| | |
| SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.2----------NOTE- --------
| |
| Not required to be met if SR 3.10.8.3 satisfied.
| |
| Perform the MODE 2 applicable SRs for LCO 3.3.2.1, According to the Function 2 of Table 3.3.2.1-1. applicable SRs SR 3.10.8.3----------NOTE- --------
| |
| Not required to be met if SR 3.10.8.2 satisfied.
| |
| Verify movement of control rods is in compliance with During control rod the approved control rod sequence for the SDM test by movement a second licensed operator or other qualified member of the technical staff.
| |
| SR 3.10.8.4 Verify no other CORE ALTERATIONS are in progress. 12-hc'-rs
| |
| -j (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 1 3.10-22 Brunswck Unt1 310-22Amendment No. O
| |
| | |
| SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.8.5 Verify each withdrawn control rod does not go to the -Eachtime the withdrawn overtravel position. control rod is withdrawn to "full out" position
| |
| _AND Prior to satisfying LCO 3.10.8.c requirement after work on control rod or CRD System that could affect coupling SR 3.10.8.6 Verify CRD charging water header pressure > 940 psig.
| |
| IIn accordance with the II Surveillance Frequency Control Program S
| |
| Brunswick Unit 1 3.10-23 Brunswck UntI 310-23Amendment No.20
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals Control Room Envelope Habitability Procqram (continued)
| |
| : e. The quantitative limits on unfiltered air inleakage into the ORE. 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 ORE 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 ORE habitability, determining ORE unfiltered inleakage, and measuring ORE pressure and assessing the ORE boundary as required by paragraphs c and d, respectively.
| |
| 41=
| |
| 5.5.14 Surveillance Frequency Control Prociram This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| : 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. Ohanges 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.
| |
| Brunswick Unit 1 5.0-17a BrunwickUnit1No. 248 I 5.-17aAmendment
| |
| | |
| BSEP 15-0101 Enclosure 4 Marked-up Technical Specification Pages - Unit 2
| |
| | |
| Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between the monitored Once within core keff and the predicted core keff is within+/-+1% Ak/k. 24 hours after reaching equilibrium conditions following startup after fuel mnovement within the reactor. pressure vessel or control rod replacement AND 14,!00MWD7 ni MODE !
| |
| I naccordance with the Surveillance Frequency ljI Control Program Brunswick Unit 2 3.1-6 Brunsick Uit 2 .1-6Amendment No. 2
| |
| | |
| Control Rod OPERABILITY 3.1.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 2.4--heurs SR 3.1.3.2 ---------- NOTE- --------
| |
| Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
| |
| 7' Insert each withdrawn control rod at least one notch 3m~
| |
| SR 3.1.3.3 Verify each control rod scram time from fully n accordance with to notch position 06 is < 7 seconds. SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.1-10 Bruswik No. 2-78 I Uit 3.-10Amendment
| |
| | |
| Controi Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify, for a representative sample, each tested control ,200-da...
| |
| rod scram time is within the limits of Table 3.1.4-1 with eumu*lati-ve reactor steam dome pressure _>800 psig. "p....÷;"- ;t SR 3.1.4.3 Verify each affected control rod scram time is within nrror to declaring the limits of Table 3.1.4-1 with any reactor steam dom* control rod pressure. OPERABLE after work on control rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram tim* is within Prior to exceeding the limits of Table 3.1.4-1 with reactor st am dome 40% RTP after fuel pressure > 800 psig. /movement within the affected core cell AND In accordance with the I*-
| |
| Surveillance Frequency Prior to exceeding Control Program 40% RTP after work on control rod or CRD System that could affect scram time Brunswick Unit 2 3.1-13 Bruswik Uit No. 264' I 3.-13Amendment
| |
| | |
| Control Rod Scram Accumulators 3.1.5 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control rod C.1 Verify all control rods Immediately upon scram accumulators associated with inoperable discovery of inoperable with reactor accumulators are fully charging water steam dome pressure inserted, header pressure
| |
| < 950 psig. < 940 psig AND C.2 Declare the associated 1 hour control rod inoperable.
| |
| D. Required Action B.1 or C.1 D.1-------NOTE----
| |
| and associated Completion Not applicable if all Time not met. inoperable control rod scram accumulators are associated with fully inserted control rods.
| |
| Manually scram the reactor. Immediately SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure is 7-d4ays
| |
| _>940 psig.,j I n accordance with the Surveillance Frequency Control Program II l Brunswick Unit 2 3.1-17 Bruswik Uit 3.-17Amendment No. 2-32
| |
| | |
| Rod Pattern Control 3.1.6 ACTIONS (continued) ________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1-------NOTE----
| |
| control rods not in Control rod may be compliance with BPWS. bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1.
| |
| Suspend withdrawal, of Immediately control rods.
| |
| AND B.2 Manually scram the reactor. 1 hour SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with BPWS. 2-e~
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.1-19 BrunwickUnit2 3.-19Amendment No. 223
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS ___________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem to 7 days inoperable. OPERABLE status.
| |
| B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable, subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate solution 24-heir-s is within the limits of Figure 3.1.7-1. I,,*
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.1-20 Bruswik Uit 3.-20Amendment No. 2438
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate solution is 24-heuf within the limits of Figure 3.1.7-2.
| |
| SR 3.1.7.3 Verify temperature of pump suction and discharge I2A-h~s piping up to the SLC injection valves is withinthl of Figure 3.1.7-2. 1/h SR 3.1.7.4 Verify continuit SR 3.1.7.5 Verify the concentration of boron in solutic 4*yt the limits of Figure 3.1.7-1.
| |
| Once within In accordance with the
| |
| * 24 hours after water Surveillance Frequency or boron is added to Control Program solution I
| |
| AND Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-2 (continued)
| |
| Brunswick Unit 2 3.1-21 BrunwickUnit2 3.-21Amendment No. 25 I
| |
| | |
| SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each pump develops a flow rate _ 41.2 gpm at a In accordance with discharge pressure Ž>1190 psig the Inservice Testing Program SR 3.1.7.7 Verify flow through one SLC subsystem from pump 21! monthc on into reactor pressure vessel STAGGERED TEST SR 3.1.7.8 Verify sodium pentaborate enrichment is > 47 atom n*ror to addition to percent B-10. SLC tank I
| |
| IIn accordance with the Surveillance Frequency Control Program f
| |
| I Brunswick Unit 2 3.1-22 Bruswik Uit 3.-22Amendment No. 2-*5
| |
| | |
| SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 -NO TE-........................
| |
| Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.
| |
| Verify each SDV vent and drain valve is open.
| |
| A SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully closed and fully open position.
| |
| m SR 3.1.8.3 Verify each SDV vent and drain valve:
| |
| 2A-months
| |
| : a. Closes in < 30 seconds after receipt of ai or simulated scram signal; and
| |
| : b. Opens when the actual or simulated signal is reset.
| |
| SIn accordanceFrequency Surveillance with the Control Program i
| |
| I Ii Brunswick Unit 2 3.1-26 Bruswik Uit 3.-26Amendment No. 2-a2
| |
| | |
| APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| LCO 3.2.1 COLR.
| |
| APPLICABILITY: THERMAL POWER Ž_23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 4 hours limits. within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion Time POWER to < 23% RTP.
| |
| not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND 2hors hrcfc iF In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 2 3.2-1 Bruswik Uit 3.-1Amendment No. 24-7-i
| |
| | |
| MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| LCO 3.2.2 specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER Ž>23% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 4 hours Iimits. limits.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND 24 hours thcreafer (continued)
| |
| In accordance with the Surveillance Frequency Control Program j'I Brunswick Unit 2 3.2-2 BrunwickUni 2 I 32-2Amendment No. 247-
| |
| | |
| LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER >Ž23% RTP.
| |
| ACTIONS_________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 4 hours limits, within limits.
| |
| B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 23% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within specified in the COLR. 12 hours after
| |
| Ž>23% RTP AND t-her-eaf te SIn accordance with the Surveillance Frequency Control Program j'I l III Brunswick Unit 2 3.2-4 Bruswik Uit 3.-4Amendment No. 2--74 I
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS
| |
| ...................................... NOTES-----------------------------..............
| |
| : 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.1 (Not used.)
| |
| SR 3.3.1.1.2 Perform CHANNEL CHECK. 24-h~eur-SR 3.3.1.1.3 ------------ NOTE----------------
| |
| Not required to be performed until 12 hours after THERMAL POWER >Ž23% RTP.
| |
| \
| |
| Adjust the average power range monitor (APRM) channels to conform to the calculated power while hi operating at > 23% RTP.
| |
| SR 3.3.1.1.4-----------NOTE--------
| |
| Not required to be performed when entering MODE2, fr~om _MO_ DE_ 1_until 12 _h~ours_ after_ entering_ MOD E_ 2 ._
| |
| Perform CHANNEL FUNCTIONAL TEST.
| |
| 1l (continued)
| |
| In accordanceFrequency Surveillance with the Control Program II Brunswick Unit 2 3.3-4 Brunwic Uni 233-4Amendment No. 24-7
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform a functional test of each automatic scram i-day contactor.
| |
| SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to withdrawing intermediate range monitor (IRM) channels overlap SRMs from the fully inserted position SR 3.3.1.1.7--------------------.......NOTE--------
| |
| Only required to be met during entry into MOI2 ro MODEl1.
| |
| Verify the IRM and APRM channels overl p.
| |
| SR 3.3.1.1.8 Calibrate the local pwranemonit rs. 2000 ,cf,,ti,,cfull power rnge .. .............
| |
| SR 3.3.1.1.9 Perform CHANNEL FUNCTION ET §o ,d...
| |
| SR 3.3.1.1.10 Calibrate the trip units. Q112- ....
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 2 3.3-5 Bruswik Uit 3.-5AmendmentNo. 2-82 I
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.11 -~~~~NOTES----- ---
| |
| : 1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
| |
| : 2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.
| |
| Perform CHANNEL FUNCTIONAL TEST.
| |
| 444dt SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.
| |
| SR 3.3.1.1.13 -~~NOTES--------- I
| |
| : 1. Neutron detectors are excluded. )
| |
| : 2. For Function 1, not required to be perf when entering MODE 2 from MODE 1 12 hours after entering MODE 2. )
| |
| : 3. For Functions 2.b and 2.f, the rec transmitters that feed the APRMs Perform CHANNEL CALIBRATION. 24, mcnthe SR 3.3.1.1.14 (Not used.)
| |
| SR 3.3.1.1.15 Perform LOGIC SYSTEM FL LA -nenth*
| |
| (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program 1
| |
| I Brunswick Unit 2 3.3-6 BrunwickUni 2 33-6Amendment No. 242,I
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)__________
| |
| SURVEILLANCE JFREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve--Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure--Low Functions are not bypassed when THERMAL POWER is _>26% RTP.
| |
| SR 3.3.1.1.17 ---------------......... NOTES- ------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. For Functions 3 and 4, the sensor response /
| |
| (
| |
| time may be assumed to be the design senso-response time./
| |
| i* l," I*lii.-I#,*,l.*Al,*
| |
| * fl*fl * .,*1* A *1.. .... I* ,,il*_ J.I V.11111tz~I§ iui 1 r ,,q!Iue ,n,.e, 4 ........
| |
| ulluri .*I(:
| |
| 1" I TEcOT rA') "Ie - . ...
| |
| I-jul I rIIrlr'r"' ~*"~-"*
| |
| outputs ,- hll-*I a-lt+r-.atc..-
| |
| STAGGERED TEST Verify the RPS RESPONSE TIME is Il SR 3.3.1.1.18 Adjust the flow control trip refe* Once within 7 days to reactor flow. after reaching equilibrium conditions following refueling outage (continued)
| |
| SIn accordanceFrequency with the l Surveillance
| |
| * Control Programl II I Brunswick Unit 2 3.3-7 Bruswik Uit 3.-7Amendment No. 24-7
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM Simulated 24-meiaths Thermal Power is > 25% and recirculation drive flow is
| |
| < 60%.
| |
| In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.3-8 BrunwickUni 2 33-8Amendment No. 24*
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS
| |
| ----------------...................... NOTE----------------------------..............
| |
| Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. 12-hc,-rs SR 3.3.1.2.2 1.
| |
| 2.
| |
| -------- NOTES---------..............
| |
| Only required to be met during CORE ALTERATIONS.
| |
| One SRM may be used to satisfy more than of the following.
| |
| 1 Verify an OPERABLE SRM detector is located in:
| |
| : a. The fueled region; 42-hcurs
| |
| : b. The core quadrant where CORE ALTERA are being performed, when the associated is included in the fueled region; and /
| |
| : c. A core quadrant adjacent to where COF*
| |
| ALTERATIONS are being performed, v4h associated SRM is included in the fuel4!d SR 3.3.1.2.3 Perform CHANNEL CHECK. . a-eu-I (continued)
| |
| Im
| |
| * Surveillance Frequency lIn accordance Control Program with the Brunswick Unit 2 3.3-14 BrunwickUnit2 No. 24 I 3.-14Amendment
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.4 -~~~~NOTES--------
| |
| : 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
| |
| : 2. Not required to be met during a core spiral off load.
| |
| Verify count rate is >_3.0 cps.
| |
| AI4P-DATI*%
| |
| SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST.
| |
| lay SR 3.3.1.2.6 -~~NOTE----
| |
| Not required to be performed until 12 hours a*
| |
| on Range 2 or below. /
| |
| Perform CHANNEL FUNCTIONAL TES]
| |
| (continued Surveillance Frequency IIn accordance Control with the Program I
| |
| Brunswick Unit 2 3.3-15 Bruswik Uit 3.-15Amendment No. 242,
| |
| | |
| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.2.7------------NOTES--------
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
| |
| Perform CHANNEL CALIBRATION. 24-m~em'ths Inaccordance with the SureilaceFrequency Control Program Brunswick Unit 2 3.3-16 BrunwickUnit2No. 24 I 3.-16Amendment
| |
| | |
| Control Rod Block Instrumentation 3.3.2,1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor Mode E.1 Suspend control rod Immediately Switch--Shutdown Position withdrawal.
| |
| channels inoperable.
| |
| AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ...................................... NOTES- - - - - - - - - - - - - - -
| |
| : 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
| |
| : 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
| |
| SURVEILLANCE FREUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. -- dy I (continued)
| |
| In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 2 3.3-20 Bruswik Uit 3.-20Amendment No.
| |
| * Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.2-----------NOTE- -------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn at _<8.75% RTP in MODE 2.
| |
| Not required to be performed until 1 hour after THERMAL POWER is*_<8.75% RTP in MODE 1.
| |
| Perform CHANNEL FUNCTIONAL TEST. -dy SR 3.3.2.1.4 Verify the RBM: 2A-ReRths
| |
| : a. Low Power Range--Upscale Function OR Intermediate Power Range--Upscale Fur ti OR High Power Range--Upscale Functic enabled (not bypassed) when APRM Si u e Thermal Power is _>29%.
| |
| : b. Intermediate Power Range--Upscale tio OR High Power Range--Upscale Fun t ni enabled (not bypassed) when APRM iua Thermal Power is _>Intermediate Po eRa Setpoint specified in the COLR.
| |
| : c. High Power Range--Upscale Func i enabled (not bypassed) when APRS lae Thermal Power is __High Power R* g etpoint specified in the COLR.
| |
| Y (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.3-21 BrunwickUnit2 No. 24-7 I 3.-21Amendment
| |
| | |
| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when THERMAL 2A-meat-s POWER is
| |
| * 8.75% RTP.
| |
| ;SH 3.3.Z. 1.f5 ------------- "U I -------------------.......
| |
| Not required to be performed until 1 hour after reactor*
| |
| mode switch is in the shutdown position.
| |
| Perform CHANNEL FUNCTIONAL TEST. 124-mGeth SR 3.3.2.1.7 ----------- NOTE-----------
| |
| Neutron detectors are excluded.
| |
| Perform CHANNEL CALIBRATION. 24-meiMhs SR 3.3.2.1.8 Verify control rod sequences input to the Prior to declaring conformance with BPWS. RWM OPERABLE following loading of sequence into RWM IIn accordanceFrequency Surveillance with the Control Program I
| |
| Brunswick Unit 2 3.3-22 BrunwickUnit2 3.-22Amendment No. 2-4 Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE------ ---------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided feedwater and main turbine high water level trip capability is maintained.
| |
| SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK. 24-heus SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The Allowable 2me*
| |
| Value shall be < 207 inches. .
| |
| SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24, mcnth*
| |
| including valve actuation. ,1 In accordance with the Surveillance Frequency Control Program
| |
| !/
| |
| Brunswick Unit 2 3.3-25 BrunwickUnit2 3.-25Amendment No. 22*
| |
| | |
| PAM Instrumentation 3.3.3.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion Time referenced in of Condition C not met. Table 3.3.3.1-1 for the channel.
| |
| E. As required by Required E.1 Be in MODE 3. 12 hours Action D.1 and referenced in Table 3.3.3.1-1.
| |
| F. As required by Required F.1 Initiate action in accordance Immediately Action D.1 and referenced in with Specification 5.6.6.
| |
| Table 3.3.3.1-1.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE------ ---------
| |
| These SRs apply to each Function in Table 3.3.3.1-1.
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. ,3-,-d....
| |
| SR 3.3.3.1.2 (Not Used.) /
| |
| I (continued)
| |
| In accordance with the Surveillance Frequency Control Program F/
| |
| ..m Brunswick Unit 2 3.3-27 Bruswck ni 2 .327Amendment No. 2 PAM Instrumentation 3.3.3.1 SURVEILLANCE REQU IREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.1.3 Perform CHANNEL CALIBRATION for each required 24-me~tthe PAM Instrumentation channel. .A I IIn accordance with the Surveillance Frequency Control Program l I Brunswick Unit 2 3.3-28 Bruswck ni 2 .328Amendment No. 26 Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Remote Shutdown Monitoring Instrumentation LCO 3.3.3.2 OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS*
| |
| ------------- NOTE Separate Condition entry is allowed for each Functih I
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable, to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 1dy instrumentation channel that is normally energized.
| |
| (continued) lIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 2 3.3-30 BrunwickUnit2 3.-30Amendment No. 2-60
| |
| | |
| Remote Shutdown Monitoring Instrumentation 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each required 2A-e*hs instrumentation channel.
| |
| In accordance with the Surveillance Frequency Control Program tl Brunswick Unit 2 3.3-31 Brunsick Uit 2 .3-31Amendment No. 2
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One Function with B.1 Restore ATWS-RPT trip 72 hours ATWS-RPT trip capability capability.
| |
| not maintained.
| |
| C. Both Functions with C.1 Restore ATWS-RPT trip 1 hour ATWS-RPT trip capability capability for one Function.
| |
| not maintained.
| |
| D. Required Action and D.1 Remove the associated 6 hours associated Completion Time recirculation pump(s) from not met. service.
| |
| O__R D.2 Be in MODE 2. 6 hours SURVEILLANCE REQUIREMENTS
| |
| ------------------------- I,.1 I--------------------------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 2A. he'-r (continued)
| |
| IIn accordance with the qi Surveillance Frequency Control Program I II Brunswick Unit 2 3.3-33 BrunwickUnit2 3.-33Amendment No. 22*
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 1~
| |
| SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST. 92-days f-SR 3.3.4.1.3 Calibrate the trip units.
| |
| SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The Allowable a4-me~ths Values shall be:
| |
| : a. Reactor Vessel Water Level--Low Level *.
| |
| Ž_101 inches; and/
| |
| : b. Reactor Vessel Pressure--High: _<114 SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL. [2A-msiths including breaker actuation.
| |
| lIn accordanceFrequency Surveillance with the Control Program I
| |
| I I
| |
| Brunswick Unit 2 3.3-34 BrunwickUnit2 3.-34Amendment No. 22-a
| |
| | |
| ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS
| |
| ---------------...................... NOTES------ ---------
| |
| : 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c; and (b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.3-40 Brunsick Uit 2 .3-40Amendment No. 2
| |
| | |
| RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.
| |
| Brunswick Unit 2 3.3-47 Brunsick Uit 2 .3-47Amendment No. 2
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES---------------
| |
| : 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions 2.c, 2.d, 3.a, 3.b, 3.e, 3.f, 3.g, 3.h, 4.a, 4.b, 4.e, 4.f, 4.g, 4.h, 4.i, 4.k, 5.a, 5.b, 5.e, 5.f, and 6.a; and (b) for up to 6 hours for all other Functions provided the associated Function maintains isolation capability.
| |
| Brunswick Unit 2 3.3-52 Brunsick Uit 2 .3-52Amendment No. 2-
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.1.8 -~~~~NOTES----- --
| |
| : 1. Radiation detectors are excluded.
| |
| : 2. The sensor response time for Functions 1 .a, 1 .c, and 1.f may be assumed to be the design sensor response time.
| |
| Verify the ISOLATION INSTRUMENTATION RESPONSE TIME is within limits. ESTAGGERED SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST. i llm9 I~l i In accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 2 3.3-53 Brunsick Uit 2 .3-53Amendment No. 2
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 hour secondary containment isolation dampers inoperable.
| |
| AND C.2.1 Place the associated 1 hour standby gas treatment (SGT) subsystem(s) in operation.
| |
| O__R C.2.2 Declare associated SGT 1 hour subsystem(s) inoperable.
| |
| SURVEILLANCE REQUIREMENTS
| |
| --------------- NOTES----- ----------
| |
| : 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability.
| |
| SURVEILLANCE .LFREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 2-ew In accordance with the I J cniud Surveillance Frequency Control Program Brunswick Unit 2 3.3-60 Amendment No. 2,34
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 Brunswick Unit 2 3.3-61 BrunwickUnit2 3.-61Amendment No. 2-*
| |
| | |
| CREV System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREV initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. -- Q2-days SR 3.3.7.1.3 Perform CHANNEL CALIBRATION. 2A-meait-h Brunswick Unit 2 3.3-64 Bruswck ni 2 .364Amendment No. 267-
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Isolate condenser vacuum 12 hours associated Completion Time pumps.
| |
| of Condition A not met.
| |
| OR O__R B.2 Isolate main steam lines. 12 hours Condenser vacuum pump isolation capability not O__RR maintained.
| |
| B.3 Be in MODE 3. 12 hours SURVEILLANCE REQUIREMENTS
| |
| ---------------- NOTE----------------
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK. 24-heur-s SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST. ~ye SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The Allc !48-months Value shall be _< 6 x background. /
| |
| (continued)
| |
| SIn accordanceFrequency Surveillance with the Control Program I
| |
| lII Brunswick Unit 2 3.3-67 Bruswck ni 2 .367Amendment No. 267-
| |
| | |
| Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 24-meaths including condenser vacuum pump trip breaker and isolation valve actuation.-/
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.3-68 Bruswck ni 2 .368Amendment No. 2-6 LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS
| |
| --------- NOTES------------------------------..............
| |
| : 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Functions maintains initiation capability for three DGs; and (b) for Function 2, the associated Function maintains OG initiation capability.
| |
| Brunswick Unit 2 3.3-70 Bruswck ni 2 .370Amendment No. 267
| |
| | |
| RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to fully insert Immediately associated Completion Time all insertable control rods in of Condition A or B not met core cells containing one or in MODE 3, 4, or 5 with any more fuel assemblies.
| |
| control rod withdrawn from a core cell containing one or more fuel assemblies.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.1 ------------ NOTE---------
| |
| Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 for
| |
| >_24 hours.
| |
| Perform CHANNEL FUNCTIONAL TEST. 48A-deys SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS _A-mePths motor generator set electric power monitoring assembly. The Allowable Values shall be:
| |
| : a. Overvoltage _<129 V.
| |
| : b. Undervoltage > 105 V.
| |
| : c. Underfrequency > 57.2 Hz.
| |
| (continued)
| |
| In accordance with the v/
| |
| Surveillance Frequency Control Program II Brunswick Unit 2 3.3-73 Bruswck ni 2 .373Amendment No. 26 RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS 24-meat-hs alternate power supply electric power monitoring assembly. The Allowable Values shall be:
| |
| : a. Overvoltage _<132 V.
| |
| : b. Undervoltage _>108 V.
| |
| : c. Underfrequency _> 57.2 Hz.
| |
| SR 3.3.8.2.4 Perform a system functional test. 24-menths SIn accordance with the Surveillance Frequency Control Program
| |
| !II Brunswick Unit 2 3.3-74 Bruswck ni 2 .374Amendment No. 267-
| |
| | |
| Recirculation Loops Operating 3.4.1 ACTIONS (continued) _________
| |
| COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met.
| |
| OR No recirculation loops in operation.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1-----------NOTE- -------
| |
| Not required to be performed until 24 hours after both recirculation loops are in operation Verify recirculation loop jet pump flow mismatch with 2"4.heurs both recirculation loops in operation: ,
| |
| : a. < 10% of rated core flow when operating at
| |
| < 75% of rated core flow; and/
| |
| : b. < 5% of rated core flow when operating at A
| |
| _ 75% of rated core flow.
| |
| SIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.4-2 BrunwickUni2 34-2Amendment No. 2 Jet Pumps 3.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 -~~~~NOTES---------
| |
| : 1. Not required to be performed until 4 hours after associated recirculation loop is in operation.
| |
| : 2. Not required to be performed until _>25% RTP.
| |
| Verify at least one of the following criteria (a or b) is satisfied for each operating recirculation loop:
| |
| 2A1k
| |
| : a. Recirculation pump flow to speed ratio differs by
| |
| _<5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by _<5% from established patterns.
| |
| : b. Each jet pump diffuser to lower plenum differential pressure differs by _<10% from that jet pump's established pattern.
| |
| H-IIn accordance with the Surveillance Frequency Control Program l l Brunswick Unit 2 3.4-4 Bruswik Uit 3.-4Amendment No. 227
| |
| | |
| S RVs 3.4.3 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.3.2-----------NOTE----................
| |
| Not required to be performed until 12 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify each re~quired SRV opens when manually2",m th actuated.
| |
| In accordance with the Surveillance Frequency Control Program i
| |
| Brunswick Unit 2 3.4-6 Brunsick Uit 2 .4-6Amendment No.23
| |
| | |
| RCS Operational LEAKAGE 3.4.4 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Pressure boundary LEAKAGE exists.
| |
| SURVEILLANCEREQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and 8hu-unidentified LEAKAGE increase are within limits. .,
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| * II Brunswick Unit 2 3.4-8 Brunsick Uit 2 No. ~
| |
| .4-8Amendment
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required primary --- hr-containment atmosphere radioactivity monitoring system.
| |
| SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of requi leakage detection instrumentation.
| |
| SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required 2A-methe leakage detection instrumentation.
| |
| m SIn accordanceFrequency with the Surveillance Control Program I!
| |
| I I
| |
| Brunswick Unit 2 3.4-11 Brunsick Uit 2 No. ~~
| |
| .4-11Amendment
| |
| | |
| RCS Specific Activity 3.4.6 ACTIONS (continued' CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours AND B.2.2.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.6.1------------NOTE----------------
| |
| Only required to be performed in MODE 1.
| |
| Verify reactor coolant DOSE EQUIVALENT I-131 specific activity is < 0.2 pCi/gm.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.4-13 Brunsick Uit 2 No. 3
| |
| .4-13Amendment
| |
| | |
| RHR Shutdown Cooling System--Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1-----------NOTE---------
| |
| Not required to be met until 2 hours after reactor steam dome pressure is less than the RHR shutdown cooling isolation pressure.
| |
| Verify one required RHR shutdown cooling subsystem 42-hc'-rs or recirculation pump is operating. l IIn accordance with the Surveillance Frequency Control Program l!
| |
| Brunswick Unit 2 3.4-16 Brunsick Uit 2 .4-16Amendment No.23
| |
| | |
| RHR Shutdown Cooling System--Cold Shutdown 3.4.8 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hour from subsystem in operation. circulating by an alternate discovery of no method. reactor coolant AN D circulation No recirculation pump in AND operation.
| |
| Once per 12 hours thereafter AND B.2 Monitor reactor coolant Once per hour temperature.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one required RHR shutdown cooling subsystem !2 hc'dr or recirculation pump is operating. "*
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.4-18 Bruswik Uit 3.-18Amendment No. 2334
| |
| | |
| RCS P/T Limits 3.4.9 ACTIONS (continued)__________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C.-----NOTE---------....C.1 Initiate action to restore Immediately Required Action C.2 shall be parameter(s) to within completed if this Condition is limits.
| |
| entered.
| |
| -------- AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for operation. MODE 2 or 3.
| |
| MODES 1, 2, and 3.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.1-----------NOTE- -------
| |
| Only required to be performed during RCS heatup and cooldown operations.
| |
| Verify:
| |
| : a. RCS pressure and RCS temperature are within /
| |
| the applicable limits specified in Figures 3.4.9-1 and 3.4.9-2; and
| |
| : b. RCS heatup and cooldown rates are < 100°F in any 1 hour period.
| |
| ~(continued)
| |
| SSurveillance Frequency SIn accordance with the Control Program Brunswick Unit 2 3.4-20 BrunwickUnit23.-20Amendment No. 22*
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.2-----------NOTE---------
| |
| Only required to be performed during ROS inservice leak and hydrostatic testing.
| |
| Verify: "*n ,-r ,,,i..
| |
| : a. RCS pressure and RCS temperature are within ,
| |
| the applicable'limits specified in Figure 3.4.9-3; 3.4.9-4, or 3.4.9-5, as applicable;
| |
| : b. RCS heatup and cooldown rates are < 30°F in any 1 hour period.
| |
| SR 3.4.9.3 Verify RCS pressure and RCS temperature are within ; nce within the criticality limits specified in Figure 3.4.9-2. j5 minutes prior to o['ntrol rod ithdraa for the purpose of achieving criticality SR 3.4.9.4---------------------NOTE---------------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the bottom head coola tOnce within temperature and the reactor pressure vessel (RPV) 30 minutes prior to coolant temperature is < 145°F. each startup of a recirculation pump (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program 1I Brunswick Unit 2 3.4-21 BrunwickUnit2 3.-21Amendment No. 2,56
| |
| | |
| RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.4.9.5-----------NOTE- -------
| |
| Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump start.
| |
| Verify the difference between the reactor coolant Once within temperature in the recirculation loop to be started and 30 minutes prior to the RPV coolant temperature is _<50°F. each startup of a recirculation pump SR 3.4.9.6-----------NOTE- -------
| |
| Only required to be performed when tensioning the reactor vessel head bolting studs.
| |
| Verify reactor vessel flange and head flange ,=-R'-mI'Wte&
| |
| temperatures are >_70°F.
| |
| SR 3.4.9.7-----------NOTE- -------
| |
| Not required to be performed until 30 minutes after ROS temperature _<80°F in MODE 4. I Verify reactor vessel flange and head flange ~0-miR~4tes temperatures are _ 70°F.
| |
| ~6 SR 3.4.9.8-----------NOTE- -------
| |
| Not required to be performed until 12 hours after RC:*
| |
| temperature < 100°F in MODE 4.
| |
| Verify reactor vessel flange and head flange /
| |
| temperatures are _>70°F.*! 1!2-he'-s In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.4-22 Bruswik Uit 3.-22Amendment No. 24?2
| |
| | |
| Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be < 1045 psig.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is _<1045 psig. 42-hc',w'r I
| |
| IIn accordance with the Surveillance Frequency Control Program II II Brunswick Unit 2 3.4-28 BrunwickUnit2 No. 25 I 3.-28Amendment
| |
| | |
| ECCS--Operating 3.5.1 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME J. Two or more low pressure J.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A or B.
| |
| O__R HPCI System and two or more required ADS valves inoperable.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.5.1 .1 Verifyj, for each ECCS injection/spray subsystem, the 4-days ..
| |
| piping is filled with water from the pump discharge A valve to the injection valve.
| |
| V (continued)
| |
| In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.5-4 BrunwickUni 2 35-4Amendment No. 247
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.2 -------- NOTE----------...............
| |
| Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) shutdown cooling isolation pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path,/
| |
| that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.5.1.5--------------------Nticspl e---------------- 4...
| |
| SR3514 Verify eahe RHRSsemrcu pupdicarge osation vas nokd Once-achsaru (contnued BuSwikUi 2... 3.5-5E Amndet o.
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.6 Verify the following ECCS pumps develop the 92-de~s specified flow rate against a system head corresponding to the specified reactor pressure. 4-SYSTEM HEAD CORRESPONDIN(
| |
| NO. OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS >4l00gpm 1 >_113psig LPCI > 14,000 gpm 2 > 20 psig SR 3.5.1.7--------------------NOTE--------- --------
| |
| Not required to be performed until 48 hours a ner reactor steam pressure is adequate to perfor the test.
| |
| Verify, with reactor pressure < 1045 and "945 psig, 24L the HPCI pump unit can develop a flow r te
| |
| >_4250 gpm against a system head cor pnigt reactor pressure.
| |
| SR 3.5.1.8------------NOTE----------------...
| |
| Not required to be performed until 4t hours aft reactor steam pressure is adequate operforrth test. *g Verify, with reactor pressure __1I psghe HPCI 244 menthc pump unit can develop a flow ra .. Ž4 0 gpm against a system head corresponding t_ e or pressure (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.5-6 Brunsick Uit 2 No. a~
| |
| .5-6Amendment
| |
| | |
| ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SU RVEI LLANCE FREQU ENCY SR 3.5.1.9---------------------NOTE---------------
| |
| Vessel injection/spray may be excluded.
| |
| Verify each ECCS injection/spray subsystem actuates 2"!. mRths on an actual or simulated automatic initiation signal.
| |
| SR 3.5.1.10--------------------NOTE----------------
| |
| Valve actuation may be excluded.
| |
| Verify the ADS actuates on an actual or simulated 24-FPcethe automatic initiation signal._.
| |
| SR 3.5.1.11-----------NOTE----------------
| |
| Not required to be performed until 12 hours after reactor steam pressure is adequate to performte test.
| |
| Verify each required ADS valve opens when anu I 4-RnthS injection/spray subsystem is within the I*ir lIn accordance with the Surveillance Frequency Control Program Y/
| |
| Brunswick Unit 2 3.5-7 Brunsick Uit 2 .5-7Amendment No. 2
| |
| | |
| ECCS--Shutdown 3.5.2 ACTIONS (continued) _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| : 0. Required Action C.2 and 0.1 Initiate action to restore Immediately associated Completion Time secondary containment to not met. OPERABLE status.
| |
| AND D.2 Initiate action to restore one Immediately standby gas treatment subsystem to OPERABLE status.
| |
| AND 0.3 Initiate action to restore Immediately isolation capability in each required secondary containment penetration flow path not isolated.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure coolant injection !2-hc'--'s (LPCI) subsystem, the suppression pool water level is
| |
| >-31 inches. 7 SIn accordance with the I cniud SurvillnceFrequency Control Program Brunswick Unit 2 3.5-9 Amendment No. 22g
| |
| | |
| ECCS-Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE jFREQUENCY Verify, for each required core spray (CS) subsystem, SR 3.5.2.2 the: 1--h\
| |
| : a. Suppression pool water level is >_-31 inches; or b.---------NOTE----------......
| |
| Only one required CS subsystem may take credit for this option during OPDRVs.
| |
| I
| |
| --------------------- I Condensate storage tank water volume is
| |
| _> 228,200 gallons.
| |
| SR 3.5.2.3 Verify, for each required ECCS injection/spray subsystem, the piping is filled with water from the l pump discharge valve to the injection valve.
| |
| SR 3.5.2.4 One LPCI subsystem may be considered OEA during --------
| |
| alignment and operation for decay-*-E-------
| |
| NOTE----..... heatre if capable of being manually realigned and not other~wise inoperable.
| |
| Verify each required ECCS injection/spray subsysln manual, power operated, and automatic valve intl securedflwin thaitisot,is in the correct position.o /A II .~.'
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.5-10 Bruswik Uit 3.-10Amendment No. 2-&
| |
| | |
| ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the specified flow rate against a system head corresponding to the specified reactor pressure. /
| |
| NO. CORRESPONDI NG OF SYSTEM HEAD G TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS >__4100gpm 1 Ž__113 psig LPCI _>9000 gpm 1 _>20 psig SIR 3.5.2.6 -NOTE--------.........----
| |
| Vessel injection/spray may be excluded. I Verify each required ECCS injection/spray subsl actuates on an actual or simulated automatic min signal.
| |
| SR 3.5.2.7 -NOTE----------......
| |
| Instrumentation response time may be assumeq the design instrumentation response time. J Verify the ECCS RESPONSE TIME for each 2"!,months ECCS injection/spray subsystem is within thq SIn accordanceFrequency Surveillance with the Control Program l
| |
| I l
| |
| Brunswick Unit 2 3.5-11 3.-11Amendment BrunwickUnit2 No. 24*
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water from ~-1-d.a...
| |
| the pump discharge valve to the injection valve.
| |
| SR 3.5.3.2 Verify each RCIC System manual, power operated, and automatic valve in the flow path, that is not lock, sealed, or otherwise secured in position, is in the 31di correct position.
| |
| SR 3.5.3.3 ----------- NOTES ...
| |
| : 1. Use of auxiliary steam for the performance o the SR is not allowed./
| |
| : 2. Not required to be performed until 24 hour.,
| |
| reactor steam pressure is adequate to per~
| |
| the test.I Verify, with reactor pressure _>945 psig and <
| |
| psig, the RCIC pump can develop a flow ratej
| |
| >_ 400 gpm against a system head correspon ct reactor pressure.
| |
| I (continued) lIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.5-13 Brunsick Uit 2 .5-13Amendment No. 2
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.4 ------------- NOTES- ------
| |
| : 1. Use of auxiliary steam for the performance of the SR is not allowed with reactor pressure
| |
| _>150 psig.
| |
| : 2. Not required to be performed until 24 hours after reactor steam pressure is adequate to perform the test.
| |
| Verify, with turbine inlet pressure _ 135 psig and < 165 24-meRmhs psig, the RCIC pump can develop a flow rate
| |
| Ž>400 gpm against a system head corresponding to an*
| |
| equivalent reactor pressure.
| |
| 4 SR 3.5.3.5 ------------ NOTE-------.......... -
| |
| Vessel injection may be excluded./
| |
| Verify the RCIC System actuates on an actual or 2"t-menthe simulated automatic initiation signal.
| |
| In accordance with the Surveillance Frequency Control Program I
| |
| h II Brunswick Unit 2 3.5-14 Brunsick Uit 2 .5-14Amendment No. 2
| |
| | |
| Primary Containment 3.6.1.1 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and leakage rate In accordance with testing, except for primary containment air lock testing, the Primary in accordance with the Primary Containment Leakage Containment Rate Testing Program. Leakage Rate Testing Program SR 3.6.1.1.2 Verify drywell to suppression chamber differential 2-eth pressure does not decrease at a rate > 0.25 inch water gauge per minute tested over a 10 minute period at an I initial differential pressure of_> 1.00 psid and
| |
| *<1.25 psid.
| |
| Brunswick Unit 2 3.6-2 Brunsick Uit 2 .6-2Amendment No. 2
| |
| | |
| Primary Containment Air Lock 3.6.1.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1-----------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 3.6.1.1.1.
| |
| Perform required primary containment air lock leakage In accordance with rate testing in accordance with the Primary the Primary Containment Leakage Rate Testing Program. Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary containment air 24m th lock can be opened at a time. I" Brunswick Unit 2 3.6-6 Amendment No. 2-2
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.6.1.3.1 -~~~~NOTES--------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCI Vs that are open under administrative controls.
| |
| Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident 31dJ conditions is closed.
| |
| p.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.6-11 Brunsick Uit 2 .6-11Amendment No. 2
| |
| | |
| PCI Vs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 1-SR 3.6.1.3.2 ------------ NOTES--------
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCI Vs that are open under administrative controls.
| |
| Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside primary MODE 2 or 3 from containment and not locked, sealed, or otherwise MODE 4 if primary secured and is required to be closed during accident containment was conditions, is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.3 Verify continuity of the traversing incore probe (TIP) .a1,da...
| |
| shear isolation valve explosive charge.
| |
| SR 3.6.1.3.4 Verify the isolation time of each power operated and In accordance each automatic PCIV, except for MSIVs, is within with the Inservice limits. Testing Program SR 3.6.1.3.5 Verify the isolation time of each MSIV is _>3 seco d In accordance with and
| |
| * 5 seconds. the Inservice Testing Program VI-
| |
| *L*UI ILII lU{E:*LI)
| |
| IIn accordanceFrequency with the Surveillance Control Program I II Brunswick Unit 2 3.6-12 Bruswik Uit 3.-12Amendment No. 24=2
| |
| | |
| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify each automatic PCIV actuates to the isolation 24-meP4hs position on an actual or simulated isolation signal.
| |
| SR 3.6.1.3.7 Verify a representative sample of reactor 24-mefPths instrumentation line EFCVs actuate to the isolation A
| |
| *position on an actual or simulated instrument line break signal.
| |
| SR 3.6.1.3.8 Remove and test the explosive squib from each sarIn accordance with isolation valve of the TIP System. the Inservice Testing Program SR 3.6.1.3.9 Verify leakage rate through each main steam lirs In accordance with
| |
| _<100 scfh and the combined leakage rate of alfu the Primary main steam lines is < 150 scfh when tested at Containment
| |
| _>25 psig. Leakage Rate Testing Program IIn accordanceFrequency Surveillance Control Program with the i
| |
| Brunswick Unit 2 3.6-13 Brnsic Uit2 .613Amendment No. 2-67
| |
| | |
| Drywell Air Temperature 3.6.1.4 3,6 CONTAINMENT SYSTEMS 3,6.1.4 Drywell Air Temperature LCO 3.6.1.4 Drywell average air temperature shall be < 150°F.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A.1 Restore drywell average air 8 hours temperature not within limit, temperature to within limit.
| |
| B. Required Action and. B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREUENCY SR 3.6.1.4.1 Verify drywell average air temperature is within limit. 2-e~
| |
| I IIn accordance with the Surveillance Frequency Control Program II II Brunswick Unit 2 3.6-14 Brunsick Uit 2 .6-14Amendment No. 2
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.5 SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify nitrogen bottle supply pressure of each nitrogen 2A.-hc'-r backup subsystem is __ 1130 psig.
| |
| SR 3.6.1.5.2----------NOTES--------
| |
| : 1. Not required to be met for vacuum breakers tht are open during Surveillances.
| |
| : 2. Not required to be met for vacuum breakers open when performing their intended functi n SR 3.6.1.5.4 Verify the full open setpoint of each vacuu brars2A-,mcthe
| |
| _<0.5 psid.
| |
| SR 3.6.1.5.5 Verify leakage rate of each nitrogen ba kp24-meRt-he SR 3.6.1.5.6 Verify the Nitrogen Backup System s e ir 21-. mcnthe actuation signal.
| |
| Surveillance Frequency Control Program Brunswick Unit 2 3.6-17 Bruswik Uit 3.-17Amendment No. 22
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 NOTE -.
| |
| Not required to be met for vacuum breakers that are open during Surveillances.
| |
| Verify each vacuum breaker is closed. 44-d~e
| |
| ~AND Within 6 hours after any discharge of steam to the
| |
| / suppression chamber from any source AND Within 12 hours following an operation that causes any of the vacuum breakers to open.
| |
| SR 3.6.1.6.2 Perform a functional test of each r breaker. /
| |
| I, AND Within 12 hours after any discharge of IIn accordanceFrequency Surveillance with the steam to the Control Program suppression chamber from the S RVs vacuum{_ .......... ___
| |
| SR 3.6.1.6.3 Verify theis full breaker open
| |
| < 0.5 setpoint of each required psid. 24 ,*Ie~he Brunswick Unit 2 3.6-19 BrunwickUnit2 No. 2-7-9 I 3.-19Amendment
| |
| | |
| Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature is within 2A-heu~s the applicable limits. AN 5 minutes when
| |
| * performing testing that adds heat to the suppression pool l
| |
| SIn accordance with the Surveillance Frequency Control Program f
| |
| l Brunswick Unit 2 3.6-22 Bruswik Uit 3.-22Amendment No. 2-32
| |
| | |
| Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be _>-31 inches and < -27 inches.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression pool 2 hours level not within limits, water level to within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE / FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits. 2Ahc'-rs SIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.6-23 Brunsick Uit 2 .6-23Amendment No. 2
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem ~-*4-d.e,,
| |
| manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise\
| |
| SR 3.6.2.3.2 Verify each RHR pump develops a flow rate / 2-deys while operating in the suppression pool cooling m de./
| |
| In accordance with the Surveillance Frequency Control Program I /I Brunswick Unit 2 3.6-25 Brunsick Uit 2 .6-25Amendment No. 2
| |
| | |
| Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LCO 3.6.3.1 The primary containment oxygen concentration shall be < 4.0 volume percent.
| |
| APPLICABILITY: MODE 1 during the time period:
| |
| : a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to
| |
| : b. 24 hours prior to a scheduled reduction of THERMAL POWER to
| |
| < 15% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment oxygen A.1 Restore oxygen 24 hours concentration not within concentration to within limit.
| |
| limit.
| |
| B. Required Action and B.1 Reduce THERMAL 8 hours associated Completion Time POWER to < 15% RTP.
| |
| not met.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.3.1.1 Verify primary containment oxygen concentration is :7-days within limits. _
| |
| Surveillance SIn accordanceFrequency with the Control Program Brunswick Unit 2 3.6-26 Brunsick Uit 2 .6-26Amendment No. 2
| |
| | |
| Secondary Containment 3.6.4.1 ACTIONS _______
| |
| COMPLETION CONDITION REQUIRED ACTION TIME C. (continued) C.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment equipment hatches 24-meFR-hs are closed and sealed. ,
| |
| SR 3.6.4.1.2 Verify one secondary containment access door is 24-m ^n'lh closed in each access opening.
| |
| SR 3.6.4.1.3 Verify each SGT subsystem can maintain > 0.25 in* 21 months on--
| |
| of vacuum water gauge in the secondary containm STAG GERI*EDI1" for 1 hour at a flow rate <*3000 cfm.
| |
| I-I#
| |
| lIn accordance with the Surveillance Frequency Control Program I
| |
| l II Brunswick Unit 2 3.6-29 Bruswik Uit 3.-29Amendment No. 2-80 i
| |
| | |
| SCIDs 3.6.4.2 ACTIONS (continued) _________________________
| |
| COMPLETION CONDITION REQUIRED ACTION TIME D. Required Action and D.1-----------NOTE--- .
| |
| associated Completion Time LCO 3.0.3 is not applicable.
| |
| of Condition A or B not met - - - -
| |
| during movement of recently irradiated fuel assemblies in Suspend movement of Immediately the secondary containment recently irradiated fuel or during OPDRVs. assemblies in the secondary containment.
| |
| AND D.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.2.1 Verify the isolation time of each automatic SCID is 2A!. menth*
| |
| within limits.
| |
| SR 3.6.4.2.2 Verify each automatic SCID actuates to the isolation 2A-imqethe position on an actual or simulated actuation signal.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.6-32 Bruswik No. 2-8Q I Uit 3.-32Amendment
| |
| | |
| SGT System 3.6.4.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two SGT subsystems E.1 -NOTE---......
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of recently irradiated fuel ..-.......... ..... ...-
| |
| assemblies in the secondary containment or during Suspend movement of Immediately OPDRVs. recently irradiated fuel assemblies in secondary containment.
| |
| AND E.2 Initiate action to suspend Immediately OPDR Vs.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SOT subsystem for >_10 continuous ~ 1dy hours with heaters operating.I*
| |
| SR 3.6.4.3.2 Perform required SOT filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). i, the VFTP SR 3.6.4.3.3 Verify each SOT subsystem actuates on an actual o simulated initiation signal.
| |
| lIn accordanceFrequency Surveillance with the Control Program III Brunswick Unit 2 3.6-35 Bruswik No. 2-00 I Uit 3.-35Amendment
| |
| | |
| RHRSW System 3.7.1 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSW manual, power operated, and v, .. *-
| |
| automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the I" correct position or can be aligned to the correct position.
| |
| * IIn accordance with the II Surveillance Frequency Control Program y
| |
| II Brunswick Unit 2 3.7-3 BrunwickUni 2 37-3Amendment No. 2-*
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in the SW pump suction bay of 24-he'-rs the intake structure is > -6 ft mean sea level.
| |
| SR 3.7.2.2 Verify the water temperature of UHS is < 90.5°F. *4hu SR 3.7.2.3 ----------- NOTE- -------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each SW System manual, power operated, automatic valve in the flow paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| /
| |
| SR 3.7.2.4
| |
| : 1. A single test at the specified Frequency will
| |
| -------- NOTE---------
| |
| satisfy this Surveillance for both units.
| |
| : 2. Isolation of flow to individual components not render SW System inoperable.
| |
| Verify automatic transfer of each DG cooling . Q2~de~Le supply from the normal SW supply to the alter supply on low DG jacket cooling water supply pressure. A (continued)
| |
| IIn accordanceFrequency Surveillance with the Control Program I
| |
| I II Brunswick Unit 2 3.7-9 BrunwickUni 2 37-9Amendment No. 240
| |
| | |
| SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.5-----------NOTE-- -------
| |
| Isolation of flow to individual components does not render SW System inoperable.
| |
| Verify each required SW System automatic component 2A-imcnths actuates on an actual or simulated initiation signal.
| |
| * I n accordance with the SSurveillance Frequency SControl Program Brunswick Unit 2 3.7-10 BrunwickUnit2 3.-10Amendment No. 2-*
| |
| | |
| CREV System 3.7.3 ACTIONS (continued)
| |
| COMPLETION CONDITION REQUIRED ACTION TIME E. Two CREV subsystems--------------NOTE-----------
| |
| inoperable during movement LCO 3.0.3 is not applicable.
| |
| of irradiated fuel assemblies- ------------
| |
| in the secondary containment, during CORE E.1 Suspend movement of Immediately ALTERATIONS, or during irradiated fuel assemblies OPDRVs. in the secondary containment.
| |
| OR AND One or more CREV subsystems inoperable due E.2 Suspend CORE Immediately to an inoperable CRE ALTERATIONS.
| |
| boundary during movement of irradiated fuel assemblies AND!
| |
| in the secondary containment, during CORE E.3 Initiate action to suspend Immediately ALTERATIONS, or during OPDRVs.
| |
| OPDRVs.
| |
| SuRvEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREV subsystem for > 15 continuous "gA- ...
| |
| minutes.
| |
| SR 3.7.3.2 Perform required CREV filter testing in accordance naccordance with with the Ventilation Filter Testing Program (VFTP). the VFTP (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.7-13 BrunwickUnit2 No. 2-7-6 I 3.-13Amendment
| |
| | |
| CREV System 3.7.3 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.7.3.3 Perform required CRE unfiltered air inleakage testing In accordance with in accordance with the Control Room Envelope the Control Room Habitability Program. Envelope Habitability Program SR 3.7.3.4 Verify each CREV subsystem actuates on an actual or 2"!, mcnths simulated initiation signal. *1 II In accordance with the Surveillance Frequency Control Program
| |
| //
| |
| I Brunswick Unit 2 3.7-14 Bruswik Uit No. 2 I 3.-14Amendment
| |
| | |
| Control Room AC System 3.7.4 ACTIONS (continued) _________________ _________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME F. Three control room AC---------NOTE- ----
| |
| subsystems inoperable LCO 3.0.3 is not applicable during movement of irradiated fuel assemblies in the secondary containment, F.1 Suspend movement of Immediately during CORE irradiated fuel assemblies in ALTERATIONS, or during the secondary containment.
| |
| O PD RVs.
| |
| AND F.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND
| |
| !F.3 Initiate actions to suspend Immediately OPDRVs.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify each control room AC subsystem has the 24-me*4-hs capability to remove the assumed heat load. ,,
| |
| Surveillance Frequency In accordance with the Control Program tJ Brunswick Unit 2 3.7-17 BrunwickUnit2 3.-17Amendment No. 2-32,
| |
| | |
| Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -NOTE--------------------.......
| |
| Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
| |
| Verify the gross gamma activity rate of the noble gases is < 243,600 IJCi/second after decay of 30 mnts Once within 4 hours afteraŽ_>50%
| |
| increase in the nominal steady 1Surveillance Frequency state fission gas Control Prga release after factoring out increases due to changes in THERMAL POWER level Brunswick Unit 2 3.7-19 Brunsick Uit 2 .7-19Amendment No. 2
| |
| | |
| Main Turbine Bypass System 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete cycle of each main turbine bypass *" '- ...
| |
| valve.
| |
| SR 3.7.6.2 Perform a system functional test. " 4Re*h SR 3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESPONSE'1 24. mc~the TIME is within limits./i IControl Program Brunswick Unit 2 3.7-21 Brunsick Uit 2 .7-21Amendment No. 2
| |
| | |
| Spent Fuel Storage Pool Water Level 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Spent Fuel Storage Pool Water Level LCO 3.7.7 The spent fuel storage pool water level shall be _>19 feet 11 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool A.1-------NOTE- --
| |
| water level not within limit. LCO 3.0.3 is not applicable.
| |
| Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the spent fuel storage pool water level is y
| |
| _>19 feet 11 inches over the top of irradiated fuel ,
| |
| assemblies racks. seated in the spent fuel storage pool =
| |
| l SIn accordance with the Surveillance Frequency Control Program
| |
| /
| |
| II Brunswick Unit 2 3.7-22 Bruswik Uit 3.-22Amendment No. 2438
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7dy availability for each offsite circuit.Ii SR 3.8.1.2------------..............NOTES----------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. A modified DG start involving idling and gradua acceleration to synchronous speed may be us(
| |
| for this SR. When modified start procedurese not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG starts from standby conditions ad ldy achieves steady state voltage > 3750 V and < 4 00 and frequency > 58.8 Hz and < 61.2 Hz.y
| |
| ... *_*(continued)
| |
| In accordance with the Surveillance Frequency Control Program
| |
| ,11 Brunswick Unit 2 3.8-7 Bruswik Uit 3.-7Amendment No. 2-2
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS_(continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.3 ---------- NOTES---------
| |
| : 1. DG loadings may include gradual loading.
| |
| : 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.
| |
| : 5. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each OG is synchronized and loaded and ~A-deys operates for __60 minutes at a load > 2800 kW and
| |
| < 3500 kW.
| |
| SR 3.8.1.4 Verify each engine mounted tank contains Ž 150 gal ofI 31-deys fuel oil. l SR 3.8.1.5 Check for and remove accumulated water from eachA engine mounted tank. / aqye SR 3.8.1.6 Verify the fuel oil transfer system operates to tr aye fuel oil from the day fuel oil storage tank to the mounted tank.
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 2 3.8-8 Bruswik Uit 3.-8Amendment No. 2-*
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.7 ---------- NOTES---------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| 4 *A H*no Verify each DG starts from standby condition and achieves, in < 10 seconds, voltage _>3750 V and frequency >_58.8 Hz, and after steady state conditions are reached, maintains voltage >_3750 V and < 4300 V and frequency > 58.8 Hz and _<61.2 Hz.
| |
| (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program
| |
| ! II Brunswick Unit 2 3.8-9 Bruswik No. 2-8 I Uit 3.-9Amendment
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUI REMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.8 -~~~~NOTES---------
| |
| : 1. SR 3.8.1.8.a shall not be performed in MODE 1 or 2 for the Unit 2 offsite circuits. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. SR 3.8.1.8.a is not required to be met ifthe unit power supply is from the preferred offsite circuit.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify: 2A -meR÷1h*.
| |
| : a. Automatic transfer capability of the unit power supply from the normal circuit to the preferred offsite circuit; and
| |
| : b. Manual transfer of the unit power supply from the preferred offsite circuit to the alternate offsite circuit.
| |
| (continued)
| |
| Surveillance Frequency lIn accordance with the Control Program Brunswick Unit 2 3.8-10 Brunsick Uit 2 .8-10Amendment No. 2
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.9 -~~~~NOTES---------
| |
| : 1. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 3 and DG 4. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. If performed with the DG synchronized with offsite power, it shall be performed at a power factor < 0.9.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG rejects a load greater than or equal to its associated core spray pump without tripping. 2AmeL h S(continued)
| |
| Surveillance Frequency SIn accordance with the Control Program Brunswick Unit 2 3.8-11 Bruswik Uit 3.-11Amendment No. 2~z*
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY 4
| |
| SR 3.8.1.10 -NOTES, A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG's automatic trips are bypassed on an 2A-mePths actual or simulated ECCS initiation signal except:
| |
| : a. Engine overspeed;
| |
| : b. Generator differential overcurrent;
| |
| : c. Low lube oil pressure;
| |
| : d. Reverse power;
| |
| : e. Loss of field; and
| |
| : f. Phase overcurrent (voltage restrained). A I
| |
| (continued)
| |
| IIn accordance with they t Control ProgramI Brunswick Unit 2 3.8-12 Brunsick Uit 2 .8-12Amendment No. 5
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.11 -~~~~NOTES---------
| |
| : 1. Momentary transients outside the load and power factor ranges do not invalidate this test.
| |
| : 2. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| Verify each DG operating at a power factor < 0.9 operates for _>60 minutes loaded to _>3500 kW and
| |
| _<3850 kW. 24 Re. h Si-i.
| |
| SR 3.8.1.12 NOTES. -.
| |
| A single test at the specified Frequency will satisfy thisj Surveillance for both units./
| |
| Verify an actual or simulated ECCS initiation signal I:2A-menthe capable of overriding the test mode feature to returr each OG to ready-to-load operation.
| |
| (continued)
| |
| Surveillance Frequency IIn accordance with the Control Program I
| |
| al Brunswick Unit 2 3.8-13 Bruswik Uit 3.-13Amendment No. 2-*
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.13----------NOTES---------
| |
| This Surveillance shall not be performed in MODE 1, 2, or 3 for the load sequence relays associated with DG 3 and DG 4. However, credit maybe taken for unplanned events that satisfy this SR.
| |
| Verify interval between each sequenced load block is 2A-.meinthe within + 10% of design interval for each load sequence relay. ,
| |
| (continued)
| |
| In accordance with the Surveillance Frequency Control Program k/
| |
| II Brunswick Unit 2 3.8-14 Bruswik Uit 3.-14Amendment No. 2-38
| |
| | |
| AC Sources--Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.14 -~~~NOTES---------
| |
| : 1. All DG starts may be preceded by an engine prelube period.
| |
| : 2. This Surveillance shall not be performed in MODE 1, 2, or 3 for DG 3 and DG 4. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify, on actual or simulated loss of offsite power signal in conjunction with an actual or simulated ECCS initiation signal:
| |
| r2A-R
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses; and
| |
| : c. DG auto-starts from standby condition and:
| |
| : 1. energizes permanently connected loads in
| |
| _<10.5 seconds,
| |
| : 2. energizes auto-connected emergency loads through load sequence relays,
| |
| : 3. maintains steady state voltage >_3750 V and _<4300 V,
| |
| : 4. maintains steady state frequency >_58.8 Hz and
| |
| * 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected emergency loads for
| |
| > 5 minutes.
| |
| In accordance with the Surveillance Frequency Control Program IJ Brunswick Unit 2 3.8-15 BrunwickUnit2 3.-15Amendment No.
| |
| * I
| |
| | |
| Diesel Fuel Oil 3.8.3 SURVEILLANCE REQUIREMENTS________
| |
| SURVEILLANCE FREQUENCY SR 3.8.3.1 For each required OG, verify: " *..
| |
| : a. The associated day fuel oil storage tank ,
| |
| contains _>22,650 gal; and
| |
| : b. The main fuel oil storage tank contains _>20,85q gal per required 0DG.
| |
| SR 3.8.3.2 Verify fuel oil properties of stored fuel oil are testedln In accordance with accordance with, and maintained within the limits , the Diesel Fuel Oil the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.3 Check day fuelforoiland tankremove and theaccumulated main fuel oilwater from storage taps*ach k..
| |
| r SIn accordance with the Surveillance Frequency Control Program m
| |
| I m
| |
| Brunswick Unit 2 3.8-22 BrunwickUnit2 3.-22Amendment No. 22* I
| |
| | |
| DC Sources--Operating 3.8.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 4. 36 hours Two or more DC electrical power subsystems inoperable.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is _>130 V on float -ey charge. /
| |
| /I SR 3.8.4.2 Verify no visible corrosion at battery terminals and de~s connectors.
| |
| ORR Verify battery connection resistance is _<23.0 pa for inter-cell connections and < 82.8 pohms for ii rack connections.
| |
| SR 3.8.4.3 Verify battery cells, cell plates, and racks show 4g-me~the visual indication of physical damage or abnorm deterioration that degrades performance. ~1 (continued)
| |
| IIn accordance with the Surveillance Frequency Control Program II I Brunswick Unit 2 3.8-24 BrunwickUnit2 3.-24Amendment No. 235
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell !8- ienths and terminal connections are coated with anti-corrosion material. \
| |
| SR 3.8.4.5 Verify each required battery charger supplies 24-~ne~the
| |
| Ž_250 amps at _>135 V for _Ž4 hours.
| |
| SR 3.8.4.6 ------------- NOTES--------
| |
| : 1. The modified performance discharge test in SR 3.8.4.7 may be performed in lieu of the service test in SR 3.8.4.6 once per 60 months.
| |
| : 2. This Surveillance shall not be performed in MODE 1 or 2 for the Unit 2 DC electrical power subsystems. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 3. A single test at the specified Frequency will satisfy this Surveillance for both units.
| |
| I Verify battery capacity is adequate to Supply, and 24-RmeRths maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.
| |
| (continued)
| |
| SIn accordance with the Surveillance Frequency Control Program I
| |
| Brunswick Unit 2 3.8-25 Brunsick Uit 2 .8-25Amendment No.23
| |
| | |
| DC Sources--Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.7 ------------ NOTES- -------
| |
| : 1. This Surveillance shall not be performed in MODE 1 or 2 for the Unit 2 DC electrical power subsystems. However, credit may be taken for unplanned events that satisfy this SR.
| |
| : 2. A single test at the specified Frequency will.
| |
| satisfy this Surveillance for both units.
| |
| Verify battery capacity is __80% of the manufacturer's rating when subjected to a performance discharge tes*.
| |
| or a modified performance discharge t.est. / 12*mon-ths we battery shows degradation or has reached 85% of the Surveillance Frequec expected life with Control Prga capacity < 100% of manufacturer's rating AND 24 months when battery has reached 85% of the expected life with capacity > 100% of manufacturer's rating Brunswick Unit 2 3.8-26 Brunsick Uit 2 .8-26Amendment No.23
| |
| | |
| Battery Cell Parameters 3.8.6 ACTIONS__________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.
| |
| B. Required Action and B.1 Declare associated battery Immediately associated Completion Time inoperable.
| |
| of Condition A not met.
| |
| O__R One or more batteries with average electrolyte temperature of the representative cells not within limits.
| |
| OR One or more batteries with one or more battery cell parameters not within Category C limits.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 -7 Category A limits.,.,
| |
| (continued)
| |
| SIn accordance with the '
| |
| Surveillance Frequency Control Program Brunswick Unit 2 3.8-31 BruswikUit No. 24=,= I 3.-31Amendment
| |
| | |
| Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY
| |
| *I.
| |
| SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 92" r-,days Category B limits.
| |
| SR 3.8.6.3 Verify average electrolyte temperature of representative cells is Ž_60°F.
| |
| I.
| |
| IIn accordanceFrequency Surveillance with the Control Program I
| |
| I I
| |
| Brunswick Unit 2 3.8-32 BrunwickUnit2 I 3.-32Amendment No. 285
| |
| | |
| Distribution Systems--Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and indicated power 7--de~ys availability to required AC and DC electrical power distribution subsystems.
| |
| SR 3.8.7.2 Verify no combination of more than two power lays conver~sion modules (consisting of either two lightii inverters or one lighting inverter and one plant uninterruptible power supply unit) are aligned to Division IIbus B. I In accordance with the Surveillance Frequency Control Program dI Brunswick Unit 2 3.8-37 BrunwickUnit2 No. 22* I 3.-37Amendment
| |
| | |
| Distribution Systems--Shutdown 3.8.8 ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to suspend Immediately operations with a potential for draining the reactor vessel.
| |
| AND A.2.4 Initiate actions to restore Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and indicated power 7-gays availability to required AC and DC electrical power distribution subsystems.
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.8-39 BrunwickUnit2 No. 23 I 3.-39Amendment
| |
| | |
| Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of the following required refueling equipment interlock inputs:
| |
| : a. All-rods-in,
| |
| : b. Refuel platform position,
| |
| : c. Refuel platform fuel grapple, fuel loaded,
| |
| : d. Fuel grapple position,
| |
| : e. Refuel platform frame-mounted hoist, fuel loaded, and
| |
| : f. Refuel platform monorail hoist, fuel loaded.
| |
| "1-In accordance with the l Surveillance Frequency Control Program Brunswick Unit 2 3.9-2 BrunwickUni 2 39-2Amendment No. 22,3,
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
| |
| ACTIONS_________________ ___
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-out A,.1 Suspend control rod Immediately interlock inoperable., withdrawal.
| |
| AND A,.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE jFREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in Refuel position. 1 cr
| |
| - I (continued)
| |
| In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.9-3 Brunsick Uit 2 .9-3Amendment No. 2
| |
| | |
| Refuel Position One-Rod-Out Interlock 3.9.2 SURVEILLANCE REQUIREMENTS (continued)________
| |
| SURVEILLANCE FREQUENCY SR 3.9.2.2------------NOTE--------
| |
| Not required to be performed until 1 hour after any control rod is withdrawn.
| |
| Perform CHANNEL FUNCTIONAL TEST. 7ea-In accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.9-4 Brunsick Uit 2 .9-4Amendment No.
| |
| | |
| Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
| |
| APPLICABILITY: When loading fuel assemblies into the core.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control rods A.1 Suspend loading fuel Immediately not fully inserted, assemblies into the core.
| |
| SURVEILLANCE REQUIREMENTS _______
| |
| SR 3.9.3.1 Verify all control rods are fully inserted ... -r In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.9-5 Brunsick Uit 2 .9-5Amendment No. 2
| |
| | |
| Control Rod OPERABILITY--Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY--Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.
| |
| APPLICABILITY: MODE 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to fully insert Immediately control rods inoperable, inoperable withdrawn control rods.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.5.1---------------------NOTE---------------
| |
| Not required to be performed until 7 days after the control rod is withdrawn.
| |
| Insert each withdrawn control rod at least one notch. -s Verify each withdrawn control rod scram acmltr SR 3.9.5.2 pressure is Ž_940 psig./
| |
| In accordance with the Surveillance Frequency Control Program l
| |
| Brunswick Unit 2 3.9-8 Brunsick Uit 2 .9-8Amendment No.23
| |
| | |
| RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 RPV water level shall be _> 23 ft above the top of irradiated fuel assemblies seated within the RPV.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies within the RPV, During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.
| |
| ACTIONS _______
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not within A.1 Suspend movement of fuel Immediately limit. assemblies and handling of control rods within the RPV.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is > 23 ft above the top of 2 or irradiated fuel assemblies seated within the RPV. , ,
| |
| SIn accordance with the Surveillance Frequency Control Program III j
| |
| II Brunswick Unit 2 3.9-9 BruswikUit 3.-9Amendment No. 24=2
| |
| | |
| RHR--High Water Levei 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is ! cr operating.
| |
| Surveillance lnaccordanceFrequency withth Brunswick Unit 2 3.9-12 Brunsick Uit 2 .9-12Amendment No. 2
| |
| | |
| RHR--Low Water Level 3.9.8 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is hc'-rs operating.
| |
| Surveillance I naccordanceFreq withuenc the Control Program Brunswick Unit 2 3.9-15 Bruswik Uit 3.-15Amendment No. 2-
| |
| | |
| Reactor Mode Switch Interlock Testing 3.10.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Place the reactor mode 1 hour switch in the shutdown position.
| |
| OR A.3.2------NOTE-- --
| |
| Only applicable in MODE 5.
| |
| Place the reactor mode 1 hour switch in the refuel position.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted in core cells !2-he,-rs containing one or more fuel assemblies. /
| |
| Brunswick Unit 2 3.10-5 Amendment No.
| |
| | |
| Single Control Rod Withdrawal--Hot Shutdown 3.10.3 SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.3.2-----------NOTE-........................
| |
| Not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d. 1 requirements.
| |
| Verify all control rods, other than the control rod being !24-.heire SR 3.10.3.3 wihrwaeful all control rods,netdVerify other than the control rod beini *2A. c-he*
| |
| In accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.10-8 Brunsick Uit 2 .10-8Amendment No.23
| |
| | |
| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 ACTIONS (continued) _________________________
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. One or more of the above B.1 Suspend withdrawal of the Immediately requirements not met with control rod and removal of the affected control rod not associated CRD.
| |
| insertable.
| |
| AND B.2.1 Initiate action to fully insert Immediately all control rods.
| |
| OR B.2.2 Initiate action to satisfy the Immediately requirements of this LCO.
| |
| SURVEILLANCEREQUIREMENTS _______
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.4.2--------------------NOTE---------------
| |
| Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c.1 requirements.
| |
| Verify all control rods, other than the control rod being 2"!.heurs withdrawn, in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| (continued) iIn accordance with the Surveillance Frequency Control Program II Brunswick Unit 2 3.10-11 Brunwic Unt 23.1-11Amendment No. 2-*
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| Single Control Rod Withdrawal--Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS (continued) _______
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| SURVEILLANCE FREQUENCY SR 3.10.4.3 Verify all control rods, other than the control rod being 2A-heur-s withdrawn, are fully inserted.
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| SR 3.10.4.4-----------NOTE-------------
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| Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.b.1 requirements.
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| Verify a control rod withdrawal block is inserted.
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| I1 IIn accordance with the Surveillance Frequency Control Program Brunswick Unit 2 3.10-12 Brunswck Unt No. ~~
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| 2 310-12Amendment
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| Single CR0 Removal--Refueling 3.10.5 ACTIONS CONDITI}}
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