1CAN031801, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425): Difference between revisions

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Arkansas Nuclear One 1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Richard L. Anderson Site Vice President 10 CFR 50.90 1CAN031801 March 12, 2018 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555
 
==SUBJECT:==
Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425)
Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51
 
==REFERENCES:==
NUREG-1430, Standard Technical Specifications Babcock and Wilcox Plants, Revision 4, April 2012 (ML12100A177)
 
==Dear Sir or Madam:==
 
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, Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the technical specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-1).
The proposed amendment would modify the ANO-1 TSs by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, Risk-Informed Technical Specification Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies. provides a description of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides documentation of the probabilistic risk assessment (PRA) technical adequacy. Attachment 3 provides the existing TS pages marked up to show the proposed change. Attachment 4 provides revised (clean) TS pages. Attachment 5 provides the proposed TS Bases changes for information only. provides a proposed No Significant Hazards Consideration, consistent with that published in the Federal Register on July 6, 2009 (74 FR 32000). Attachment 7 provides a cross reference table that correlates ANO-1 TS surveillance requirement numbers to the NUREG-1430 (Reference) TS surveillance requirement numbers.
 
1CAN031801 Page 2 of 3 Entergy requests approval of the proposed license amendment by April 1, 2019, with the amendment being implemented within 90 days.
No new regulatory commitments are made in this submittal.
In accordance with 10 CFR 50.91, Notice for public comment; State consultation, a copy of this application, with attachments, is being provided to designated Arkansas State Official.
If you should have any questions regarding this submittal, please contact Stephenie Pyle, Manager, Regulatory Assurance, at 479.858.4704.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 12, 2018.
Sincerely, ORIGINAL SIGNED BY RICHARD L. ANDERSON RLA/dbb Attachments:
: 1. Description and Assessment
: 2. Documentation of PRA Technical Adequacy
: 3. Proposed Technical Specification Changes (markup)
: 4. Revised Technical Specification Pages
: 5. Proposed Technical Specification Bases Changes (Information only)
: 6. Proposed No Significant Hazards Consideration
: 7. Arkansas Nuclear One, Unit 1 (ANO-1) TS Surveillance Requirements (SRs) to NUREG-1430 SRs Cross-Reference
 
1CAN031801 Page 3 of 3 cc:  Mr. Kriss M. Kennedy Regional Administrator U. S. Nuclear Regulatory Commission RGN-IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas Wengert MS O-08B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205
 
ATTACHMENT 1 1CAN031801 DESCRIPTION AND ASSESSMENT to 1CAN031801 Page 1 of 5 1.0    Description The proposed amendment would modify the Arkansas Nuclear One, Unit 1 (ANO-1), Technical Specifications (TSs) 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 - RITSTF Initiative 5b. Additionally, the change would add a new program, the Surveillance Frequency Control Program (SFCP), to TS Section 5.5, Programs and Manuals.
The changes are consistent with NRC approved Industry/TSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ML090850642). The Federal Register Notice published on July 6, 2009 (74 FR 31996) announced the availability of this TS improvement.
2.0    Assessment 2.1    Applicability of Published Safety Evaluation Entergy Operations, Inc. (Entergy) has reviewed the safety evaluation provided in Federal Register Notice 74 FR 31996, dated July 6, 2009. This review included the NRC staffs model safety evaluation (SE), TSTF-425, Revision 3, and the requirements specified in Nuclear Energy Institute (NEI) 04-10, Revision 1 (ML071360456). includes Entergys documentation regarding the ANO-1 probabilistic risk assessment (PRA) technical adequacy consistent with the requirements of Regulatory Guide 1.200, Revision 2 (ML090410014), Section 4.2, and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.
Entergy has concluded that the justifications presented in the TSTF proposal and the model SE prepared by the NRC staff are applicable to ANO-1 and justify this amendment to incorporate the changes to the ANO-1 TSs.
2.2    Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3; however, Entergy proposes variations or deviations from TSTF-425, as identified below, which include differing TS surveillance numbers.
: 1. ANO-1 Surveillance Requirements (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 SE dated July 6, 2009 (74 FR 32001).
: 2. The NUREG 1430 TSTF-425 markups add the new Surveillance Frequency Control Program as TS 5.5.18. ANO-1 is adopting this new program as TS 5.5.8, which is currently unused. This is an administrative deviation from TSTF-425 with no impact on the NRC's model SE dated July 6, 2009 (74 FR 32001).
to 1CAN031801 Page 2 of 5
: 3. For NUREG-1430 SRs not contained in the ANO-1 TS, the corresponding mark-ups included in TSTF-425 for these SRs are not applicable to ANO-1. This is an administrative deviation from TSTF-425 with no impact on the NRC's model SE dated July 6, 2009 (74 FR 32001).
: 4. Periodic frequencies associated with ANO-1 TSs 5.5.2 and 5.5.13 are included in the scope of this amendment that are not identified for relocation in TSTF-425, Revision 3.
The first sentence of TS 5.5.2.b is revised as follows (deleted text in strikeout and added text in italics):
Integrated leak test requirements for each system at least once per 18 monthsa frequency in accordance with the Surveillance Frequency Control Program.
TS 5.5.13.c is revised as follows (deleted text in strikeout and added text in italics):
Total particulate concentration of the fuel oil is  10 mg/l when tested every 31 days based on ASTM D-2276, Method A-2 or A-3 at a frequency in accordance with the Surveillance Frequency Control Program; Entergy has determined that the relocation of the periodic frequencies associated with these specifications are consistent with the intent of TSTF-425, Revision 3, and with the NRCs model SE dated July 6, 2009 (74 FR 32001). The subject TS Section 5.5 frequencies are periodic frequencies and do not meet the scope exclusion criteria identified in Section 1.0, "Introduction," of the model SE. These changes are similar to the SR frequency relocation (i.e., TS 6.5.17.d frequency) that the NRC approved for Waterford Unit 3 in License Amendment 249 (ML16159A419) as described in Item 7 herein.
In accordance with TSTF-425, changes to the frequencies for these surveillances would be controlled under the SFCP. The SFCP provides the necessary administrative controls to require that surveillances related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Changes to frequencies in the SFCP would be evaluated using the NRC approved methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1.
: 5. NRC letter dated April 14, 2010 (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:
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
This statement only applies to frequencies that have been changed in accordance with the SFCP and does not apply to frequencies that are relocated but not changed.
Consistent with NUREG-1430, Revision 4 (ML12100A177), Entergy has replaced the TSTF-425 TS Bases Insert #2 with the following variations based on the context of the specific Bases:
to 1CAN031801 Page 3 of 5 Individual SR Bases -
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Multiple SR Bases -
The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
SR Bases containing additional frequencies not controlled in the SFCP -
The periodic Surveillance Frequency[ies] is[are] controlled under the Surveillance Frequency Control Program.
: 6. Due to the relocation of SR frequencies and replacing the frequencies with the statement In accordance with the Surveillance Frequency Control Program, there are multiple SRs that moved to the next page as identified in the markup of the TS pages (Attachment 3). The following new pages are added because of SRs moving to the next page:
Page 3.3.1-5      (SR 3.3.1.5 moved to Page 3.3.1-4 and Table 3.3.1-1 moved to new Page 3.3.1-5)
Page 3.11-4      (SR Notes moved to Page 3.3.11-3 and Table 3.3.11-1 moved to new Page 3.3.11-4)
Page 3.6.6-2      (SRs 3.6.6.3 and 3.6.6.4 moved to new Page 3.6.6-2)
Page 3.8.1-7      (SR 3.8.1.9 moved to new Page 3.8.1-7)
Page 3.8.6-4      (SR 3.8.6.6 moved to new Page 3.8.6-4)
These changes (SRs moving to the next page) are administrative changes with no impact on the NRC's model SE dated July 6, 2009 (74 FR 32001).
: 7. TS 5.5.5.d is revised as follows (deleted text in strikeout and added text in italics):
Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASISin accordance with the Surveillance Frequency Control Program. The results shall be trended and used as part of the 18 month assessment of the CRE boundary assessment specified in TS 5.5.5.c.
TSTF-425 includes the relocation of the frequency for NUREG 1430, SR 3.7.10.4, associated with verifying one Control Room Emergency Ventilation System (CREVS) train can maintain a positive pressure relative to adjacent area(s). This SR was revised under TSTF-448, Control Room Habitability, to perform control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope to 1CAN031801 Page 4 of 5 Habitability Program. The requirement to perform the relative pressure surveillance was included in the new NUREG 1430, TS 5.5.18, "Control Room Envelope (CRE)
Habitability Program," as TS 5.5.18.d. ANO-1 adopted TSTF-448 in Amendment 239 dated October 2009 (ML082540799), designating the Control Room Envelope Habitability Program as TS 5.5.5 with the subject surveillance requirement as TS 5.5.5.d.
Therefore, the frequency change identified for NUREG-1430 SR 3.7.10.4 in TSTF-425 is being adopted as the ANO-1 TS 5.5.5.d frequency. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model SE dated July 6, 2009 (74 FR 31996). In addition, on July 26, 2016, the NRC approved a similar SR frequency relocation (i.e., TS 6.5.17.d frequency) in the Waterford Unit 3 TSTF-425 License Amendment 249 (ML16159A419).
: 8. Entergy proposes to relocate surveillance frequencies with periodicities different from those in TSTF-425 (e.g., SR 3.3.1.2). These differences have been approved by the NRC in prior ANO-1 amendment requests and are an administrative deviation from TSTF-425 with no impact on the NRC staff's model SE dated July 6, 2009 (74 FR 31996).
Entergy proposes to relocate surveillance frequencies, except those that reference other approved programs, that are purely event-driven, are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs, or are related to specific conditions. Entergy considers the differences listed herein to be minor variations or deviations of the type permitted by TSTF-425. provides a cross-reference between ANO-1 TS SRs and the NUREG-1430 SRs included in TSTF-425. This attachment includes a summary description of the referenced TSTF-425/ANO-1 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:
NUREG-1430 SRs included in TSTF-425 and corresponding ANO-1 TS SRs with plant-specific surveillance numbers; NUREG-1430 SRs not included in TSTF-425 that meet TSTF criteria for frequency relocation and corresponding ANO-1 TS surveillances with plant-specific surveillance numbers, as applicable; NUREG-1430 SRs included in TSTF-425 that are not contained in the ANO-1 TS; and ANO-1 plant-specific TS surveillances that meet TSTF criteria for frequency relocation but are not contained in either NUREG-1430 or markups in TSTF-425.
Inclusion of Attachment 7 is provided to assist the NRC staffs review of the proposed amendment and has no impact on the NRC staffs model SE dated July 6, 2009 (74 FR 32001).
to 1CAN031801 Page 5 of 5 3.0    Regulatory Analysis 3.1    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). Entergy has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the Federal Register notice is applicable to ANO-1.
3.2    No Significant Hazards Consideration Entergy Operations, Inc. (Entergy) has reviewed the proposed no significant hazards consideration determination (NSHC) published in Federal Register 74 FR 32000, dated July 6, 2009. Entergy has concluded that the proposed NSHC presented in the Federal Register notice is applicable to Arkansas Nuclear One, Unit 1 and is provided as Attachment 6 to this amendment request, which satisfies the requirements of 10 CFR 50.91(a).
3.3    Precedent Relocation of surveillance frequencies to a licensee controlled program was approved for multiple licensees including; Cooper Nuclear Station per License Amendment No. 258 issued on March 31, 2017 (ML17061A050), D.C. Cook Units 1 and 2 per License Amendment Nos. 334 and 316, respectively, issued on March 31, 2017 (ML17045A150), and Brunswick Units 1 and 2 per License Amendment Nos. 276 and 304, respectively, issued on May 24, 2017 (ML17096A129).
3.4    Conclusion Based on the considerations discussed herein, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
4.0    Environmental Consideration Entergy has reviewed the environmental consideration included in the NRC's model safety evaluation published in the Federal Register on July 6, 2009 (74 FR 32006). Entergy has concluded that the NRC's findings presented therein are applicable to Arkansas Nuclear One, Unit 1, and the determination is hereby incorporated by reference for this application.
 
ATTACHMENT 2 1CAN031801 DOCUMENTATION OF PRA TECHNICAL ADEQUACY to 1CAN031801 Page 1 of 102 DOCUMENTATION OF PRA TECHNICAL ADEQUACY TABLE OF CONTENTS
: 1. PURPOSE ............................................................................................................................. 2
: 2. SCOPE .................................................................................................................................. 2 2.1  Surveillance Frequency Change Process ................................................................... 3 2.2  Technical Adequacy of a PRA..................................................................................... 4
: 3. ANO-1 PRA TECHNICAL ADEQUACY ................................................................................. 5 3.1  Discussion ................................................................................................................... 5 3.2  ANO-1 Internal Events and Internal Flooding PRA Model .......................................... 5 3.2.1. Plant Changes Not Yet Incorporated ............................................................. 5 3.2.2. Peer Review Facts and Observations (F&Os) ............................................... 9 3.2.3. Consistency with Applicable PRA Standards ............................................... 10 3.3  ANO-1 Fire PRA Model ............................................................................................. 48 3.3.1. Plant Changes Not Yet Incorporated ........................................................... 48 3.3.2. Peer Review Facts and Observations .......................................................... 48 3.3.3. Consistency with Applicable PRA Standards ............................................... 49 3.4  Identification of Key Assumptions ........................................................................... 101 3.5  External Events and Shutdown Considerations ...................................................... 101
: 4. CONCLUSIONS ................................................................................................................ 101
: 5. REFERENCES .................................................................................................................. 102 LIST OF TABLES Table 1    Summary of Plant Changes Not Yet Incorporated in the ANO-1 PRA ........................ 7 Table 2    List of Finding F&Os Against the ANO-1 Internal Events and Internal Flooding Models ........................................................................................................ 11 Table 3    List of SRs Assessed as CC-I in the ANO-1 Internal Events PRA Model ................. 45 Table 4    List of Finding F&Os against the ANO-1 Fire PRA Model ......................................... 50 Table 5    List of SRs Assessed as CC-1 in the ANO-1 Fire PRA Model .................................. 94 to 1CAN031801 Page 2 of 102
: 1. PURPOSE The purpose of this report is to document the technical adequacy of the Arkansas Nuclear One, Unit 1 (ANO-1) Probabilistic Risk Assessment (PRA) model to support the implementation of the Surveillance Frequency Control Program (SFCP), also referred to as Technical Specifications Initiative 5b (Reference 1). ANO-1 intends to follow the guidance provided in NEI 04-10, Revision 1 (Reference 2), in evaluating proposed surveillance test interval (STI) changes (also referred as surveillance frequency changes).
: 2. SCOPE As explained in NEI 04-10, the Technical Specifications Initiative 5b uses a risk-informed, performance based approach for establishment of the surveillance frequencies, where PRA methods are used to determine the risk impact of the revised intervals. The PRA technical adequacy is addressed through NRC Regulatory Guide (RG) 1.200 (Reference 3), which references the ASME/ANS PRA standard, RA-Sa-2009 (Reference 4), for internal events at power. Risk impacts associated with fire, seismic, external events and shutdown activities may be considered quantitatively or qualitatively.
NEI 04-10 guidance includes the five key safety principles described in RG 1.174 (Reference 5),
which are followed as part of this risk-informed Technical Specification Interval change program.
The five key safety principles are:
: 1. Change meets current regulations unless it is explicitly related to a requested exemption or rule change
: 2. Change is consistent with defense-in-depth philosophy
: 3. Maintain sufficient safety margins
: 4. Proposed increases in core damage frequency (CDF) or risk are small and consistent with the Commissions Safety Goal Policy Statement
: 5. Use performance-measurement strategies to monitor the change The ANO-1 PRA model Revision 5p0, as developed for the Equipment Out of Service (EOOS) and Mitigating System Performance Index (MSPI) applications, is the current model of record used at ANO-1 for at-power, internal events. This model and its technical content was constructed and documented to meet the ASME/ANS PRA standard (Reference 4).
The ANO-1 fire PRA model update was completed in 2016. This model was constructed to meet the requirements of NUREG/CR-6850 (References 9 and 10). The PRA model quantification methodology used at Entergy Operations, Inc. (Entergy), nuclear sites is common and well-known to the industry.
Entergys approach for maintaining, updating and documenting the PRA models at all Entergy nuclear sites is controlled in the fleet procedures. These procedures are consistent with the guidance of the ASME/ANS PRA standard (Reference 4). The procedural process is comprehensive and detailed, which in turn provides the basis for establishing and maintaining the technical adequacy of the models, as well as ensuring the models reflect the as-built, as-operated plant configuration of the sites. In addition, self-assessments and independent peer reviews are also utilized by Entergy, which reassure the confidence in the approach and overall adequacy of the models against the recognized industry standards and methodologies.
to 1CAN031801 Page 3 of 102 Sections 2.1 and 2.2 describe the general change process and PRA adequacy requirements, respectively, required to support the Initiative 5b. Section 3 documents the technical adequacy of the ANO-1 PRA model specifically.
2.1    Surveillance Frequency Change Process NEI 04-10 describes the required steps to be followed to adjust an STI. A summary is presented below.
Once the STI requiring adjustment is selected, NRC regulatory commitments are collected and reviewed. If any prohibitive commitments are identified, such are examined to determine if the commitment can be changed. If there are no prohibitive commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision proceeds. If a regulatory commitment exists and the commitment change process does not permit the change, then the STI revision is not implemented (NEI 04-10, Steps 0-4 (Reference 2)).
The PRA technical adequacy is evaluated using guidance from RG 1.200 (Reference 3).
The RG addresses the need to evaluate important assumptions that relate to key modeling uncertainties (such as reactor coolant pump seal models, common cause failure methods, success path determinations, human reliability assumptions, etc.).
Further, the RG addresses the need to evaluate parameter uncertainties and demonstrate that calculated risk metrics (i.e., CDF and large early release frequency (LERF)) represent mean values. The identified gaps to Capability Category (CC) II requirements from the endorsed PRA standards in the RG and the identified key sources of uncertainty serve as inputs to identifying appropriate sensitivity cases (NEI 04-10, Step 5 (Reference 2)).
Select the revised STI value and revise any changes to the test strategy (NEI 04-10, Step 6 (Reference 2))
Qualitative considerations or qualitative analyses are developed for the STI revision.
Qualitative considerations include surveillance test and performance history, past industry and plant-specific experience, impact on defense-in-depth protection, among other considerations (NEI 04-10, Step 7 (Reference 2))
Perform quantitative and/or qualitative PRA assessments. Steps 8 through 12 in NEI 04-10 provide details regarding the use of PRA for evaluating the STI. The use of the PRA includes: determining if the structures, systems and components (SSCs) in question are modeled in the PRA, whether the SSCs or operator actions can be modeled (and make changes to the model if possible) or not, perform qualitative assessments as needed, evaluate total and cumulative effect on CDF and LERF, and perform sensitivity studies as needed.
The results and proposed STI changes are documented and summarized for consideration by the Integrated Decision-making Panel (IDP). The IDP is usually comprised of the site Maintenance Rule expert panel, a surveillance test coordinator, and a subject matter expert. The IDP approves or rejects the STI changes (with the possibility of adjustments if applicable). If the IDP approves the STI changes, these are documented and implemented. The IDP is also responsible for reviewing the performance monitoring results and providing feedback, if the STI changes, once implemented, result in unsatisfactory performance (NEI 04-10, Steps 16-20 (Reference 2)).
to 1CAN031801 Page 4 of 102 2.2    Technical Adequacy of a PRA As previously discussed, NEI 04-10 endorses the guidance of the NRC Regulatory Guide 1.200 (Reference 3) for the PRA technical adequacy determination. For the purposes of this report, Section 4.2 of RG 1.200 is used in support of Initiative 5b licensee applications. It is important to note that the scope of Initiative 5b applications is broad, and PRA assessments needed for each application vary from application to application. The following requirements are noted in Section 4.2 as necessary to demonstrate that the technical adequacy of the PRA is of sufficient quality to support the application submittal:
: 1. To address the need for the PRA model to represent the as-designed or as-built, as-operated plant.
: 2. Identification of permanent plant changes (such as design or operational practices) that have an impact on those SSCs modeled in the PRA, but have not been incorporated in the baseline PRA model. If a plant change has not been incorporated in the PRA, the licensee provides a justification of why the change does not impact the PRA results used to support the application. This justification should be in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application decision were not adversely impacted (remained the same).
: 3. Documentation that the parts of the PRA required to produce the results used in the decision are performed consistently with the standard as endorsed in the appendices of the RG. If a requirement of the standard (as endorsed in the appendix to the RG) has not been met, the licensee is to provide a justification of why it is acceptable that the requirement has not been met. This justification should be in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application were not impacted (remained the same).
: 4. A summary of the risk assessment methodology used to assess the risk of the application, including how the base PRA model was modified to appropriately model the risk impact of the application and results (note that this is the same as that required in the application-specific regulatory guides).
: 5. Identification of the key assumptions and approximations relevant to the results used in the decision-making process. Also, include the peer reviewers assessment of those assumptions. These assessments provide information to the NRC staff in their determination of whether the use of these assumptions and approximations is appropriate for the application, or whether sensitivity studies performed to support the decision are appropriate.
: 6. A discussion of the resolution of the peer review (or self-assessment, for peer reviews performed using the criteria in NEI 00-02) facts and observations that are applicable to the parts of the PRA required for the application. This discussion should take the following forms:
a discussion of how the PRA model has been changed, a justification in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application decision were not adversely impacted (remained the same) by the particular issue.
to 1CAN031801 Page 5 of 102
: 7. The standards or peer review process documents may recognize different capability categories or grades that are related to level of detail, degree of plant specificity, and degree of realism. The licensees documentation is to identify the use of the parts of the PRA that conform to capability categories or grades lower than deemed required for the given application (Section 1-3 of ASME/ANS RA-Sa-2009).
This PRA technical adequacy report addresses the quality of the PRA to support relocation of STI frequencies to a licensee controlled document. There are no STI changes proposed for this Initiative 5b application. Items 3 and 4, above, are addressed when preparing an STI change request and are, therefore, not covered in this report. The remaining items above are discussed in Section 3.
: 3. ANO-1 PRA TECHNICAL ADEQUACY 3.1    Discussion The ANO-1 PRA models are controlled in accordance with Entergy procedures consistent with the requirements provided in the ASME/ANS PRA Standard, as previously stated in Section 2.
Entergy procedures define the process to be followed to implement scheduled and interim PRA model updates and to control the PRA model files. In addition, the procedure also defines the process for identifying, tracking, and implementing model changes, and for identifying and tracking model improvements or potential issues that may affect the model. Model changes that are identified are tracked via model change requests (MCRs), which are entered in the Entergy MCR database.
Periodic PRA model updates are typically performed at least once every four years, with the option of extending the frequency for up to two years, such that the total update period does not exceed six years. Extensions are justified showing that the PRA model continues to adequately represent the as-built, as-operated plant, and must be approved by management.
The ANO-1 PRA model 5p0 was approved in 2016. The internal flood model upgrade was developed in 2016, underwent a focused-scope peer review in early 2017, and unresolved facts and observations (F&Os) are currently being addressed. Both models follow the guidelines of RG 1.200. Section 3.2 discusses the requirements in RG 1.200 to demonstrate PRA technical adequacy, as applicable to the current ANO-1 internal events and internal flooding models.
An ANO-1 fire PRA model (Reference 10) update was completed in 2016 and was developed in accordance with NUREG/CR-6850. Section 3.3 discusses the requirements in RG 1.200 to demonstrate PRA technical adequacy, as applicable to the current ANO-1 fire PRA model.
3.2    ANO-1 Internal Events and Internal Flooding PRA Model 3.2.1    Plant Changes Not Yet Incorporated As discussed in Section 3.1, an MCR database tracks PRA issues or improvements identified by PRA personnel. The MCR database includes the identification of plant changes that could impact the PRA model.
to 1CAN031801 Page 6 of 102 As part of the PRA evaluation for each STI change request, sensitivity cases are expected to be explored for areas of uncertainty associated with unresolved items (peer review Findings for ASME/ANS PRA Standard CC II or plant changes) that would impact the results of the STI change evaluation, prior to presenting the results of the risk analysis to the IDP.
A review of the MCR database for ANO-1 identified several plant changes not yet implemented which may potentially impact the PRA. These are listed in Table 1, along with the engineering changes (ECs) and corresponding MCR numbers, as well as the potential impact on the PRA model.
to 1CAN031801 Page 7 of 102 Table 1 Summary of Plant Changes Not Yet Incorporated in the ANO-1 PRA MCR        EC or Description of Change                                  Importance to Application Number    Procedure Currently ANO-1 recirculation lines for BWST through SFP purification loop are not        The modification is intended seismically qualified. EC-3069 incorporates two valves CV-1438 and CV-1441 into line      for seismic considerations.
that will get a close signal on ESAS channel 1 and 2 respectively. If a seismic event      The potential impact will be A1-3050    EC-3069 occurs, operations will be required to close these valves per the procedure. The current  addressed by STI change estimation is that it will take the operations approximately 30 minutes to take the action evaluations performed in to close the valves.                                                                      accordance with the SFCP.
The modifications have been The NFPA-805 circuit modification described in EC41875 prevents spurious operation        incorporated in the fire PRA of the valves following fire impact to the associated circuits regardless of where the    model, but the MCR is not cable is damaged. However, the FPRA was only adjusted to remove the fire impact to        yet closed. Therefore, this A1-4837    EC 41875  the valves in fire zone 129-F. The impact in other fire zones and fire areas has not      plant change has been been removed from the FPRA.                                                                incorporated, and has no The circuits being modified are for AOVs CV-1052, CV-4400, CV-7401, CV-7402,              impact in STI change CV-7403, and CV-7404 as well as MOVs CV-1053, CV-4446, CV-5611, and CV-5612.              evaluations performed in accordance with the SFCP.
The modification and updated data has been incorporated in the fire PRA During the FPRA model revision (re-baseline) for the NFPA-805 LAR RAI responses,          model, but the MCR is not update the random failure probability for the new DC supplies to bound the modification    yet closed. Therefore, this A1-4862    EC 44500 described in EC44500. Also update the random failure probabilities assumed for the        plant change has been new AFW pump if necessary to reflect that modification.                                    incorporated, and has no impact in STI change evaluations performed in accordance with the SFCP.
to 1CAN031801 Page 8 of 102 MCR        EC or Description of Change                                    Importance to Application Number    Procedure B5b actions/equipment can potentially be credited in the model to provide an alternate feedwater and RWT/BWST makeup from the B5b portable fire pump. By having an additional pump to EFW/AFW for long term scenarios can provide some benefit for          The modification involves those scenarios.                                                                          the retrieval of additional equipment from storage and SDS-008  Also, procedure SDS-008 Change 002 contains the following statement that could add        alignment for mitigation of A1-4933    1202.008  additional actions in the PRA: IF Unit 2 has AC power, THEN consider performing          certain accident scenarios.
Security Event (1203.048), Attachment G, Cross-Tying 4160V Buses Between Units.          The potential impact will be 1203.048 Blackout Procedure 1202.008 Rev. 15 provides a reference to 1203.048 for supplying        addressed by STI change power from Unit 2 to Unit 1 via 2A9.                                                      evaluations performed in accordance with the SFCP.
Once procedure 1203.048 is reviewed, there may be additional actions that can provide enhancement to the PRA model.
EC 44044 and child ECs 46389, 46390, and 46391 make up the electrical portion of ANO-1s BDBEE (Beyond-Design Basis External Event) plan. The scope of this project        The modification involves includes installing new breakers into the existing Load Center B-5 and Motor Control      the retrieval of additional Centers B-55, B-61, and B-65, and the permanent raceway, cables, and termination          equipment from storage and EC 44044  panels required to use them during a BDBEE. Portable diesel generators will be stored    alignment for mitigation of A1-5051    and child outside of the protected area (Ref. EC-44045 ANO Flex Storage Building) at separate    certain accident scenarios.
ECs    locations to assure availability of at least one post BDBEE. A portable generator will be The potential impact will be moved into a position at the common PASS (Post Accident Sampling System) building        addressed by STI change on the west side of the auxiliary building between the ANO-1 BWST (Borated Water          evaluations performed in Storage Tank) and the ANO-2 RWT (Refueling Water Tank). Once at this location it will    accordance with the SFCP.
be connected to the permanently installed infrastructure previously described.
Reviewing procedure 1106.006 Rev. 100 (dated 1/5/2017), a change was made in section 8.3 for EFW pump suction transfer to transfer to Service Water instead of the    The modification involves CST. This change removes the allowance to transfer to the CST. This change was            use of alternate system based on the SW supply being judged as the preferred for transfer vice CST T-41 since    alignments for mitigation of it is the safety-related water supply and it is currently unknown when vortexing might    certain accident scenarios.
A1-5821    1106.006 occur when aligned to the CST. However, the current revision (dated 4/6/2016) of          The potential impact will be procedure 1203.012K Rev. 48, still maintains the CST (T-41) as an option with T-41B      addressed by STI change empty when P7A/P7B suction pressure is low, but references 1106.006 that does not        evaluations performed in credit T-41 in the EFW pump suction transfer section. The QCST T-41B, CST T-41,          accordance with the SFCP.
and Service Water are credited in the PRA to provide flow to EFW.
to 1CAN031801 Page 9 of 102 3.2.2    Peer Review Facts and Observations (F&Os)
The ANO-1 PRA model has undergone several peer reviews and self-assessments which document the model quality and identify any areas with potential for improvement. The following assessments for PRA quality have been performed and documented for the ANO-1 model:
An industry peer review of the ANO-1 probabilistic safety assessment (PSA) model Revision 2p2 was conducted by the Babcock and Wilcox Owners Group in 2002 (Reference 6). The peer review concluded that there were several areas where the ANO-1 model needed improvement. The ANO-1 PSA model update Rev 3p0 completed in August 2006 addressed the Finding-level F&Os from this peer review.
In preparation for ANO-1s transition to National Fire Protection Association (NFPA) 805 standard, a gap assessment of the ANO-1 PSA 3p0 internal events PRA model was completed. The gaps impacting the fire PRA were closed to meet the NFPA transition schedule. The ANO-1 Internal Events PSA model was updated (Rev 4p0 completed in 2009) to meet the RG 1.200, Revision 1, standards.
In August 2009, a peer review of the Rev 4p0 ANO-1 PSA Model was performed and documented in the Peer Review Report (Reference 7). This peer review documented eighty-six (86) new F&Os including forty-three (43) Findings, forty-two (42) Suggestions, and one (1) Best Practice. Most of the findings pertained to documentation issues. The conclusion of the review was that the ANO-1 PRA substantially met the ASME PRA standard at CC II or better, except for those LERF SRs reviewed against CC-I.
In September 2017, a self-assessment of all ANO-1 PRA models (internal events, external, and shutdown) was performed. Several gaps were identified and are being tracked in the Entergys Paperless Condition Reporting System (PCRS).
The ANO-1 internal events model revision 5p0 was approved in 2016. This is the current PRA model as stated in Section 2. The peer review findings and the associated resolutions (with the exception of the 12 Findings superseded by the internal flooding focused scope peer review described below), as well as the remaining unresolved findings related to the internal events PRA model, are documented in the MCR database and are presented in Table 2.
The current ANO-1 internal flooding model was developed in 2016, and a focused scope peer review was completed in February 2017 against the current ASME/ANS PRA standard and RG 1.200. The results are detailed in report ENTGANO150-REPT-001 (ENERCON report)
(Reference 8). The results of the assessment show that approximately 90% of the internal flooding Supporting Requirements (SRs) were met at the CC II level. Twenty-four (24) F&Os were issued during this peer review, including twelve (12) Findings and twelve (12)
Suggestions. The F&Os were documented in the MCR database to be resolved. The unresolved F&Os related to internal flooding, as documented in the MCR database, are judged to have minimal or no impact on the overall results. The Findings related to the internal flooding PRA model are also reported in Table 2. An update to the current internal flooding model is in progress to address the unresolved findings and to incorporate model refinements.
to 1CAN031801 Page 10 of 102 3.2.3    Consistency with Applicable PRA Standards The ANO-1 PRA model revision 5p0 was assessed to meet the ASME/ANS PRA standard (Reference 4) CC-II of the SRs, except where noted in Table 3. Current Entergy PRA documentation includes an individual self-assessment that documents how each high-level requirement (HLR) and SR are met. A gap in documentation related to this individual self-assessment was identified and documented in MCR A1-5584 (missing table from Human Reliability Analysis and Quantification packages). This documentation issue is expected to be resolved as part of the model update.
The latest full-scope peer review for ANO-1 was conducted in August 2009 using RG 1.200, Rev 2. Since then, model revision 5p0 was completed which ensured all the significant F&Os from the peer review were addressed. All the F&Os are captured and documented in the MCR database. A search of ANO-1 MCRs related to peer review F&Os was performed. Finding-level F&O MCRs related to the internal event and internal flooding models are listed in Table 2, along with their disposition/resolution, if resolved, and the impact on the application.
to 1CAN031801 Page 11 of 102 Table 2 List of Finding F&Os against the ANO-1 Internal Events and Internal Flooding Models MCR                  Applicable                                                                                          Importance to Status                              Finding/Observation                            Disposition Number                    SR(s)                                                                                            Application Reviewed PRA-A1-01-001S02 Rev 1:              Section 5.4 of the Quantification      This issue was Method specific limitations and features that notebook (PSA-ANO1-01-QU ANO-1        resolved and could impact results are not identified.      Integration and Quantification Work    therefore has no Package) documents the 'Limitations    impact on STI A1-3873    Resolved      QU-B1    Possible Resolution: Identify and document    of the Quantification Methodology      change evaluations the limitations and features of the          Results which includes the method      performed in methodology that could impact the PRA        specific limitations that could impact accordance with the results.                                      results (consistent with SR QU-B1). SFCP.
to 1CAN031801 Page 12 of 102 MCR                Applicable                                                                                    Importance to Status                        Finding/Observation                          Disposition Number                SR(s)                                                                                        Application Discussion was added to Section 4.0 (SYSTEM WALKDOWNS AND INTERVIEWS) of the main body of PRA-A1-01-001S11, Rev 1., Section 4.0      the system notebook engineering documents the plant walkdowns and system    report that describes the process engineer discussions for each system. The  that ensures the PRA model reflects system engineer discussion is part of the  the ABAO plant. Namely, the PRA latter. There is no indication in the      model has been used in day-to-day walkdown/ discussion documentation that    plant operations (via EOOS Monitor) the modeling was verified to represent the  and other processes, such as as-built, as-operated plant.                Maintenance Rule and other plant This issue was Possible Resolution: Although it is        applications, for over 15 years. Any resolved and acknowledged that system walkdowns and      errors identified through these therefore has no discussions with system engineers have      applications are documented in impact on STI A1-3876    Resolved  SY-A4    been conducted and that spatial and        either the Condition Reporting change evaluations environmental hazards were identified; the  system or the Model Change performed in documentation of these activities does not  Request database. Also, accordance with the convey that the model indeed does          EN-DC-151 requires review of plant SFCP.
represent the as-built/as-operated plant. and procedure changes to ensure Perform/document additional walkdowns      that changes that affect the PRA are and/or discussions focusing on confirmation addressed as part of the model that the model represents the as-built,    update. Section 2.1.6 (System as-operated plant, or document the          Engineering Interviews) was added satisfaction of this requirement if the    to each notebook that refers back to existing walkdowns and/or discussions have  the main body discussion and the already accomplished this goal.            individual system engineer interviews that are currently documented in the walkdown checklists.
to 1CAN031801 Page 13 of 102 MCR                Applicable                                                                                        Importance to Status                        Finding/Observation                            Disposition Number                SR(s)                                                                                            Application Reviewed PRA-A1-01-001S03-EC13903:            Additional discussions including      This issue was There is no purposeful description of the    phenomenological conditions have      resolved and phenomenological conditions associated        been documented in the updated        therefore has no AS-B3,    with each accident sequence, as required      ANO-1 ATWS notebook (PSA-              impact on STI A1-3881    Resolved  SY-A22,  by the supporting requirement.                ANO1-01-AS-01). The updated            change evaluations SY-B14 Possible Resolution: Include a description    report contains phenomenological      performed in of the phenomenological conditions for each  discussion in the introduction section accordance with the sequence.                                    to each sequence category.            SFCP.
The quantification approach which includes    (Section 4.6 of the Quantification    This issue was modularization of the IE fault trees          notebook (PSA-ANO1-01-QU ANO-1        resolved and precludes calculation of importances for      Integration and Quantification Work    therefore has no events within the modules.                    Package) documents the application    impact on STI A1-3884    Resolved  QU-D6 of 'modules'. The second table in      change evaluations Possible Resolution: Provide discussion or    the section lists the significant      performed in tabulation of significant contributors to CDF contributors to each IE fault tree    accordance with the from IEs as well as from mitigating systems. module (along with % contribution). SFCP.
Updated report PSA-A1-ANO                                                                              QU-01 (Sensitivity and Uncertainty) identifies many possible sensitivity cases and quantified most of these This uncertainty 'characterization' has not                                          This issue was cases and qualitatively assessed been performed.                                                                      resolved and those that could not be quantified.
therefore has no Possible Resolution: Perform a                The examination of uncertainty in impact on STI A1-3885    Resolved  QU-E4    characterization of the sources of            PSA-A1-ANO-01-QU-01 is change evaluations uncertainty. It is recommended that the      consistent with EPRI 1016737/
performed in EPRI 1016737/NUREG-1855 approach be          NUREG 1855 methodology.
accordance with the applied.                                      Parametric uncertainty is calculated SFCP.
for CDF and LERF results with uncertainty distributions and results for mean, 5%, 50%, 95% standard deviation provided (CDF and LERF).
 
Attachment 2 to 1CAN031801 Page 14 of 102 MCR              Applicable                                                                                          Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                              Application The PSA-A1-ANO-01-QU-01 appendix evaluated potential A1-3885                                                                      uncertainty from many of the model (continued)                                                                  elements and qualitatively discusses the impact or why it is not a covered for each.
Appendix R and Appendix S in the PSA-A1-ANO-01-QU document Reviewed PRA-A1-01-001 Rev. 1 and            (ANO-1 Integration and EN-NE-G-014. The guideline instructs that    Quantification Work Package) reviews include a comparison of the basic    provide a comparison of update 5p0 This issue was event risk importances and system            results to the previous revision 4p0 resolved and importances in the current model to the      results. This includes comparison of therefore has no previous model. This appears to address      the top cutsets (and details on why impact on STI A1-3886    Resolved  QU-D7    the BE review question but there is no        items have gone up or down) -
change evaluations explicit discussion of the BE review.        Appendix R. Appendix S provides performed in the results by Sequence including a Possible Resolution: It would be helpful for                                          accordance with the comparison of the % contribution of the BE importances to be tabulated to                                                SFCP.
each sequence for both the old and demonstrate that the BEs had been            updated models. The table also reviewed.                                    provides insights into what has changed to impact the change in results.
Reviewed PRA-A1-01-001S12, Revision 1.
Section 2.1 identifies several limitations of                                        This issue was the applicability. However, the noted        Section 2.1 of the updated LERF        resolved and limitations do not address technical          analysis, PSA-ANO1-01-LE, Rev. 0,      therefore has no limitations that might impact the use in      details the limitations including      impact on STI A1-3890    Resolved  LE-G5 applications.                                technical limitations and the potential change evaluations for them to impact use in              performed in Possible Resolution: Document the            applications.                          accordance with the limitations of the technical aspects of the                                          SFCP.
analysis.
to 1CAN031801 Page 15 of 102 MCR                Applicable                                                                                          Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                                Application Some basic events (e.g., XMP119BBAF) applied in the calculation of IE frequencies  The initiating event fault trees developed for plant-specific fault trees have  (IEFTs) have been revised              Additional used calculation method 3 in CAFTA. The        extensively during the Revision 5      information on the use of calculation method 3 (1-e-t)          update. All events in the IEFT logic    IEFT development produces a probability (always < 1) rather    have the correct calculation type and  was added to the than a frequency which can be greater          the fault tree result is a frequency    system notebooks than 1. Calculation method 1 (t) in CAFTA    (yearly frequency). The Instrument      for the model should be used for those basic events          Air (IA), Power Conversion System      revision 5 update.
whose result is intended to be a frequency    (PCS), Service Water (SW), and          This is a of failure, not a probability of failure.      Intermediate Cooling Water (ICW)        documentation A1-3893  Unresolved  IE-D1 are the systems with IEFT logic and    issue and additional A discussion of the use of this calculation each system notebook has an            discussion of method is not provided, although, during attachment documenting the IEFT        flooding results is discussion of this issue, the PRA staff development (including how event        not expected to indicated that the limitations of the selected data is developed for each modeled      impact STI approach were understood.
event). The quantification notebook    evaluations Possible Resolution: Provide a description    (PSA-ANO1-01-QU) also documents        performed in of the approach taken for calculation of the  how the IEFTs are quantified and        accordance with the basic event values within support system      how the frequencies are linked to the  SFCP.
initiating event fault trees, and include the  model.
limitations of the approach.
Initiating event frequencies are not                                                  This issue was The results in the Quantification calculated in the manner required by the                                              resolved and notebook (PSA-ANO1-01-QU ANO-1 IE-C5. The IE units are in critical years                                              therefore has no Integration and Quantification Work versus reactor (calendar) years.                                                      impact on STI A1-3894    Resolved    IE-C5                                                  Package) are presented in both per change evaluations Possible Resolution: Calculate the            reactor critical year and per reactor performed in frequencies in the units specified by SR      year. Also ensured IE frequencies accordance with the IE-C5.                                        are calculated per IE-C5.
SFCP.
to 1CAN031801 Page 16 of 102 MCR                Applicable                                                                                      Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                            Application Section 5.3 of PRA-A1-01-001S06, Revision 2 identifies those initiating events The initiating event fault trees that are quantified by means of a plant      (IEFTs) have been revised specific fault tree.                                                              The IEFTs were extensively during the Revision 5 revised during the Appendices C, D, E, and F provide            update. The base model fault tree revision 5 update to additional detail on each of the 4 modeled    logic is the starting point for the be more detailed initiating events.                            IEFTs. The revised models are and meet the PRA more thorough than past models and Per Appendices E and F, the PSA logic                                              standard. A very IE-C8,                                                fully meet the PRA Standard A1-3897  Unresolved            model is used as the starting point for the                                        small change to the QU-D6                                                  (including element IE-D1). IEFT IE model; however, a number of modeling                                            results is expected.
development is documented in simplifications are made as identified in                                          Impact of this Attachments to the individual system Appendix C. These simplifications may                                              finding is expected notebooks Instrument Air (IA), Power cause the model to fall out of compliance                                          to be assessed in a Conversion System (PCS), Service with the SY requirements.                                                          case-by-case STI Water (SW), and Intermediate evaluations.
Possible Resolution: Use the system fault    Cooling Water (ICW) system tree with necessary data (exposure time)      notebooks (PSA-ANO1-01-SY_R0) changes to evaluate IEs.
to 1CAN031801 Page 17 of 102 MCR                Applicable                                                                                      Importance to Status                        Finding/Observation                          Disposition Number                SR(s)                                                                                          Application Logic uses direct linkage of CDF sequences into LERF (system level dependencies are accounted for).
Any LERF specific HEPs contained in the fault tree logic were developed using the same HRA process as other human events. Actions that are included as part of the split Direct linkage of the CDF sequences into  fractions used in LERF have no the LERF tree assures that system level    dependency with actions in the CDF dependencies (e.g., power and cooling      model, or are not credited. For water) can be accounted for.              example, the action for the Operator to use the ADV to minimize MSSV        This issue was No HRA for human actions is provided.                                            resolved and demands is assumed to fail.
Possible Resolution: Develop human error                                          therefore has no LE-A4,                                              The action to bump the pumps to probabilities using the ANO-1 selected HRA                                        impact on STI A1-3902    Resolved  LE-C7,                                              clear the RCS loop seals may process.                                                                          change evaluations LE-E1                                              actually be detrimental, so assuming  performed in Ideally, the human and hardware            a high probability of success is      accordance with the components of Split Fraction for No        conservative. If the operators do not  SFCP.
Intentional or Unintentional RCS Depress  perform this action or fail at the pre I-SGTR for Non-SBOs should be          action, the LERF is actually separated. A JHEP analysis can then be    decreased. Thus, dependency of included in the result.                    this action with existing HFEs does not make sense.
Recovery of offsite power prior to vessel failure for LERF is modeled in a similar way as recovery of offsite power prior to core damage for CDF.
Any dependencies are already accounted for in the Recovery Rules file.
to 1CAN031801 Page 18 of 102 MCR                Applicable                                                                                        Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                            Application The use of generic and plant specific data and the application of reasonable 'priors' has been addressed in the updated data Reviewed PRA-A1-01-001S05_EC15022.
analysis for the revision 5 model Section 3.1.5 provides guidance on update. The updated documentation checking results for those components          and application of generic data where no plant failures have occurred. If      (given no relevant plant specific the check (operating hours > 0.5*MTTF) failures) meets the requirements of  This issue was fails, the generic data is used. The intent is Standard element DA-D4 (same          resolved and to guard against undue influence of not        element name/number for old and      therefore has no statistically significant plant data.
updated Standard). All posterior      impact on STI A1-3903    Resolved  DA-D4 Guidance to check results is provided in      distributions judged to be reasonable change evaluations CE-P-05.07, Rev. 01. Some posterior            through both vendor and Entergy      performed in distributions do not appear reasonable (e.g.,  review. PSA-ANO-01-DA-01              accordance with the RYT P1, T7F D1)                                Revision 0 contains the relevant      SFCP.
Possible Resolution: Provide additional        analysis on plant specific data and the application of Bayesian updating.
examples in the guidance of what constitutes "unreasonable" to improve the      The outputs from the Bayesian confidence of detection and correction.        update (excel sheets) were reviewed to ensure the results are reasonable.
The software was benchmarked against known Bayesian problems to ensure correct software operation.
to 1CAN031801 Page 19 of 102 MCR                Applicable                                                                                        Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                            Application PSA-ANO1-01-IE, Rev 0 (ANO-1 Initiating Events Analysis Work Package), Appendix D Table 14 was revised to map NUREG/CR-5750 IE Category G4 (Stuck Open:
Pressurizer PORV) to initiator
                                                                            %IORV. Previously, this IE category was incorrectly mapped to initiator
                                                                            %T2 in Table H-2 of PRA-A1                                Reviewed cutset file A1_R4_1e-13_rec_M        001S06, Revision 2. This change is and fault tree A1R4P0EOS and PRA-A1      consistent with Possible Resolution 001, Revision 1, Section 6.4. This            provided by the F&O.                  This issue was observation result from comparisons to                                              resolved and similar plants.                              The %M frequency should be            therefore has no QU-D4,                                                  increased by 4.89E-06/rcry            impact on STI A1-3905    Resolved  IE-A1,  Spurious opening of SRV evaluation needs      (0.1*4.89E-05/rcry), i.e., from 5.07E- change evaluations IE-B3    to be reexamined.                            05/rcry to 5.56E-05/rcry, to account  performed in Possible Resolution: Revisit the              for a spurious opening of an SRV to    accordance with the classification of this event as cc. Update    a full open position (10% of the      SFCP.
the initiating event analysis with the proper time). No change in the value of classification of this event.                %S, currently 5.77E-04/rcry. Given that it was too late in the ANO-1 Model 5p00 update to revise the value of %M a sensitivity analysis was included in the 5p00 update to show the associated risk increase is very small (~0.3% increase in the baseline case CDF). The value for
                                                                            %M will be revised in the Rev. 6 update.
to 1CAN031801 Page 20 of 102 MCR                Applicable                                                                                      Importance to Status                        Finding/Observation                            Disposition Number                SR(s)                                                                                          Application Reviewed PRA-A1-01-001S12, Revision 1 and PRA-A1-01-001 Revision 1. No documentation of such a review is present    Review of the LERF results were in these documents.                          performed at various points during    This issue was Sensitivity studies are performed for        the model development process.        resolved and important inputs to the analysis.            Specifically, LERF cutsets were      therefore has no included in the package for the      impact on STI A1-3906    Resolved  LE-F2    Reviewed PRA-A1-001-S02, Appendix J.        Expert Panel review meeting and      change evaluations No indication that the expert panel reviewed LERF results were discussed in the    performed in the LERF results.                            slides. Insights from the meeting    accordance with the Possible Resolution: Include a review of    were incorporated into both the final SFCP.
the LERF contributors as part of the expert  Level 1 and Level 2 models.
panel review. Document this review in the expert panel report.
The ANO-1 PRA Level 1 Success MAAP 4.0.5 provides detailed core damage    Criteria Notebook/ PRA-A1                                sequences.                                  001S014, Rev. 0 provides the success criteria and systems needed PRA-A1-01-0015S014-EC14882 for the ANO-1 PRA. The MAAP Section 4.0 A of the Level 1 Success model is not recommended for the      This issue was Criteria Notebook credits the use of MAAP analysis of LLOCA during the early    resolved and for a LLOCA blowdown phase. Based on SC-A2,                                                blowdown phase and reflooding due    therefore has no current MAAP guidance, MAAP should not SC-B3,                                                to the over-prediction of water      impact on STI A1-3907    Resolved            be used to model the blowdown and reflood SC-B4,                                                retained in the primary system. As a  change evaluations stages of a LLOCA. DBA codes should be AS-A9                                                result, MAAP analyses were not        performed in used in this case. Following reflood, MAAP used to model LLOCA during this      accordance with the can be used for the remainder of the phase of the accident sequence.      SFCP.
LLOCA.
Instead, a design basis accident Possible Resolution: Use a DBA code to      code was used to support the determine LLOCA success criteria during      LLOCA analysis and success criteria the blowdown and reflood stage.              during this phase of the accident sequence.
 
Attachment 2 to 1CAN031801 Page 21 of 102 MCR              Applicable                                                          Importance to Status            Finding/Observation                Disposition Number              SR(s)                                                              Application A LLOCA analysis was completed for the purpose of defining the limiting design requirements for the LPI flow (ANO-1 Large Break Loss of Coolant Accident, Rev. 3). The LLOCA analysis is based on the Acceptance Criteria for ECCS submitted as part of the original ANO-1 licensing basis and represented as a double ended cold leg split at the pump discharge with an area equivalent to twice the cross-sectional area of the pipe (8.55 ft2).
The LLOCA analysis for the LPI with limiting design requirements will be updated (XA) or replace (U) the A1-3907 basis document currently shown the (continued)
SC notebook.
For the LLOCA including RCS Inventory Control (U) success criteria, a MAAP run was used to demonstrate CFT success criteria meets the acceptance criteria. The MAAP model used for this analysis included the limiting LPI design requirements as determined from the ANO-1 LLOCA analysis. The MAAP model was acceptable for use in evaluating the RCS Inventory Control (U), because the success criteria were associated with restoring water levels in the RCS well after core reflood was established.
 
Attachment 2 to 1CAN031801 Page 22 of 102 MCR              Applicable                                                          Importance to Status            Finding/Observation                Disposition Number              SR(s)                                                              Application MAAP does not have limitations within these regions of the accident sequence analysis.
Updated Success Criteria Notebook (PSA-ANO1-01-AS):
Section 2.0 - Computer codes were reviewed and updated to reflect codes used including LLOCA analysis.
A1-3907                                                  Section 4.0 - 1.) Updated (continued) basis documents in A.1. and A.2; 2.) Add a note to A.1 to clarify that the limiting LPI flow requirements in the MAAP Case #10 (AX2) are consistent with the ANO-1 LLOCA analysis.
Section 7.0 - Replaced reference #4 with updated ANO-1 LLOCA analysis.
to 1CAN031801 Page 23 of 102 MCR                Applicable                                                                                      Importance to Status                          Finding/Observation                          Disposition Number                SR(s)                                                                                          Application Discussion of potential modeling uncertainty issues is included in the Sensitivity and Uncertainty document. PSA-A1-ANO-01-QU-01 (Sensitivity and Uncertainty) identified many possible sensitivity cases and quantified most of these SC-C3, cases and qualitatively assessed QU-E1, There is no discussion of identification of those that could not be quantified. This issue was AS-C3, issues related to modeling uncertainty. The examination of uncertainty in    resolved and DA-E3, PSA-A1-ANO-01-QU-01 is                therefore has no HR-I3,  Possible Resolution: Provide the            consistent with EPRI 1016737/        impact on STI A1-3909    Resolved  LE-F3,  identification of sources of modeling      NUREG 1855 methodology.              change evaluations SC-C3,    uncertainty. It is recommended that the    Parametric uncertainty is calculated  performed in SY-C3,    process described in EPRI                  for CDF and LERF results with        accordance with the LE-G4,    1016737/NUREG-1855 be incorporated.        uncertainty distributions and results SFCP.
QU-E4, for mean, 5%, 50%, 95% standard QU-F4 deviation provided (CDF and LERF).
The PSA-A1-ANO-01-QU-01 appendix evaluated potential uncertainty from most of the model elements and qualitatively discusses the impact or why it is not a covered for each.
to 1CAN031801 Page 24 of 102 MCR                Applicable                                                                                        Importance to Status                        Finding/Observation                            Disposition Number                SR(s)                                                                                            Application There is no evidence that surveillance tests have been evaluated to determine if portions of the tests or sub-elements have                                          This issue was additional successes that should or should                                          resolved and not be counted when estimating operational                                          therefore has no Plant specific failure rates were demands.                                                                            impact on STI A1-3910    Resolved  DA-C10                                                  developed based on operator logs change evaluations Possible Resolution: Perform an                which includes surveillance tests.
performed in assessment of the sub-elements of all                                                accordance with the surveillance tests to obtain accurate                                                SFCP.
operational demands to be used in the PRA data.
The primary source of failure data is the Maintenance Rule Database. This database was used to screen the component failures to determine if the Maintenance Rule Functional Failures were also PRA-relevant failures. It is suggested that an additional source of data be                                                This issue was reviewed to determine if a failure may have                                          resolved and occurred that did not result in a Mrule                                              therefore has no Functional Failure. In addition, a            The operator logs used for plant data DA-C4,                                                                                        impact on STI A1-3911    Resolved            suggestion is provided to include a            is attached to the Data Notebook DA-E2                                                                                        change evaluations discussion or tabular display of those        (PSA-ANO1-01-DA-01).
performed in failures that are excluded from the data.                                            accordance with the Possible Resolution: Perform a scrub or                                              SFCP.
review of EPIX, Condition Reports, Issue Reports, and/or plant specific LERs to determine if there are any additional failures that should be considered in the PRA Data Update to supplement the Maintenance Rule Database functional failures.
to 1CAN031801 Page 25 of 102 MCR                Applicable                                                                                      Importance to Status                        Finding/Observation                            Disposition Number                SR(s)                                                                                          Application Interviews with knowledgeable plant personnel are documented in various locations in the PRA Data Analysis in various assumptions and as sources of data to estimate run times, demands, etc. It is not clear that Outage UA has been excluded from the Mrule Data since some                                            This issue was Mrule functions may be outage related.                                            resolved and There is no consideration for alignment of                                        therefore has no the AAC during a dual unit SBO. In addition    The outage times were considered in impact on STI A1-3913    Resolved  DA-C13    it appears that the SU2 transformer could      the development of plant specific change evaluations be credited to support both trains on Unit 2,  data.
performed in however it is assumed to be aligned to                                            accordance with the Unit 1.                                                                            SFCP.
Possible Resolution: Model the AAC unavailability to support either unit in the event that both or either unit requires it for LOSP mitigation and provide a documented basis for the flag alignment settings used for the SU2 transformer.
to 1CAN031801 Page 26 of 102 MCR                Applicable                                                                                              Importance to Status                          Finding/Observation                              Disposition Number                SR(s)                                                                                                  Application Section 4.0 of PRA-A1-01-001S06,                Review of the loss of Service Water Revision 2 describes the process used to        initiator removed the loss of SW identify initiating events. This process        Train A and Train B initiators based on the fact that the cross-tie valves considered generic events as well as initiating events modeled in similar plants      are normally open. The %T8 initiator (TMI and Oconee). It was noted that the          is based on a total loss of SW and thus addresses this MCR.
loss of Service Water initiator is relatively low significance in the ANO-1 PRA model.        As discussed in the SW system            This issue was
                              %T8, %T9, and %T10 are identified as Loss        notebook, Attachment 1, the success      resolved and of Running SW, Loss of SW Loop 1 and            criteria for the loss of SW initiator is therefore has no A1-3916                        Loss of SW Loop 2 respectively. There are        based on one of three SW pumps.          impact on STI and      Resolved  IE-A1    no initiators that represent a total loss of all (See PSA-ANO1-01-SY,                    change evaluations A1-5377                        3 SW Pumps (including common cause of            Appendix 13, Attachment 1).              performed in all three to fail). An initiator for Loss of    The following is the updated            accordance with the Lake is included in the model but that does      information from MCR A1-5377.            SFCP.
not account for SW pump failures. In similar NSSS designs, the loss of SW is a                Added Loss of SW Loop initiators significant contributor to CDF. The process      %T9 and %T10 and updated total is further prescribed in Fleet engineering      loss of SW initiator %T8. Also guide EN-NE-G-006.                              corrected several SW modeling errors from other Rev. 5 changes, Possible Resolution: Include the total loss      primarily associated with system flow of SW in the ANO-1 PRA.                          diversion.
to 1CAN031801 Page 27 of 102 MCR                Applicable                                                                                  Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                      Application Reviewed PRA-A1-01-001S03-EC13903:
The development of the Event Trees appears to be consistent with the plant design/operation. An isolated model error was found related to placement of an Human Action to cooldown and depressurize during a Steam Generator                                          This issue was Tube Rupture. The error results in no          The model includes HFE        resolved and account for HEP probability to fail to initiate LHF1RCSDHP, which has been    therefore has no the cooldown process high enough up in the      renamed as LPI-HFC-FO-RCSDH,  impact on STI A1-3917    Resolved  AS-A5 SGTR event tree. A HEP should be place          which addresses Failure to    change evaluations near the top gate to yield simple sequences    Cooldown RCS and Isolate Break performed in where an SGTR occurs, no equipment              (SGTR) with DHR System.        accordance with the failures occur but the operators fail to                                      SFCP.
cooldown and depressurize.
Possible Resolution: Add a HEP to the SGTR sequence model high enough in the model logic to verify that OPS successfully initiates the SGTR cooldown.
to 1CAN031801 Page 28 of 102 MCR                Applicable                                                                                          Importance to Status                          Finding/Observation                          Disposition Number                SR(s)                                                                                              Application The sentences in Table 3 of PRA-A1-01-001S06 describing operator response for the Steam Line Break initiator were deleted from the text since they are not relevant to the basis for the screening of this initiator.
The opening paragraph of the Spurious HPI Actuation initiating event screening previously In Table 3 of PA-A1-001S06, Rev. 2, where    suggested that the reason the screened potential initiating events are    initiator was screened was based on considered, two human recovery actions                                                This issue was the assumption of successful are used to justify not modeling an event as                                          resolved and operator response to terminate the an initiator (i.e., Steam line break and HPI                                          therefore has no injection prior to occurrence of an actuation). However, no justification (i.e.,                                          impact on STI A1-3919    Resolved  IE-C3                                                overpressure trip. However, this is training or procedures) was provided.                                                  change evaluations not the case. A detailed discussion performed in Possible Resolution: Provide the            of how this event was screened on a accordance with the appropriate training documents or            probabilistic basis (including operator SFCP.
procedures showing these particular human    response failure) is provided in the actions.                                    table. The misleading sentences in the opening paragraph were deleted.
Steam line break is not screened on the basis of a recovery action, but rather on the following basis included in Table 3: "The probability of a steam leak occurring in this small section of piping near the ICW pumps, provides a qualitative basis for concluding that an initiator for this failure would not be significant."
 
Attachment 2 to 1CAN031801 Page 29 of 102 MCR              Applicable                                                                                            Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                                Application Spurious HPI Actuation screening involves operator action HHF1NOTT7P - Failure to Terminate Spurious HPI Prior to Reactor Trip, A1-3919 which is now quantified in the EPRI (continued)
HRA Calculator including citations of procedures and timing assumptions that provide justification for the action.
The screening performed in Table 3 of PRA-A1-01-001S06, Rev. 2 generally follows the conditions specified in SR IE-C6.
However, the conditions in this SR are not    The only initiating event screening explicitly involved, i.e., the 10e-8 frequency                                          This issue was that relies on probabilistic analysis is was used with an OOM argument to                                                        resolved and for the Spurious HPI Actuation event.
discount the initiator, not criteria (a).                                              therefore has no A detailed discussion of how this impact on STI A1-3920    Resolved  IE-C6    Possible Resolution: As the ASME/ANS          event was screened on a change evaluations PRA Standard is currently written, the        probabilistic basis (including the performed in screening criteria in IE-C6 needs to be        basis for meeting IE-C6 numerical accordance with the used. The calculation for spurious HPI        screening criteria) is provided in SFCP.
actuation needs to be checked. Add            Table 3 of PRA-A1-01-001S06.
spurious HPI actuation due to spurious ESAS actuation as an additional initiating event.
to 1CAN031801 Page 30 of 102 MCR                Applicable                                                                                        Importance to Status                          Finding/Observation                            Disposition Number                SR(s)                                                                                            Application The data analysis update has The ANO-1 PRA Peer Review road map included a review of Maintenance indicated that the pre-initiator events have                                        This issue was Rule Functional Failures and no new been reviewed against plant-specific                                                resolved and pre-initiator events have been failures.                                                                            therefore has no identified.
Possible Resolution: To be assessed at                                              impact on STI A1-3923    Resolved  HR-C2                                                  In addition, condition reports were CC II/III for this SR, a list of existing                                            change evaluations reviewed for latent human errors.      performed in pre-initiator events at ANO-1 needs to be The attached list of condition reports accordance with the prepared, and it needs to be compared to reviewed is included in the attached  SFCP.
the list in Table 2. Events not appearing in link. None of the events would Table 2 would need to be added.
impact the HRA evaluations.
No documentation exists describing comparisons with similar plants or other                                            This issue was plant specific codes to check the                                                    resolved and reasonableness and acceptability of the      A section has been added to the        therefore has no results of the thermal/hydraulic, structural, Success Criteria Notebook (PSA-        impact on STI A1-3928    Resolved  SC-B5 or other supporting engineering bases that    ANO1-01-AS-03) to address the          change evaluations support the success criteria.                comparison of results.                performed in accordance with the Possible Resolution: Perform a comparison                                            SFCP.
with other plants and document.
to 1CAN031801 Page 31 of 102 MCR                Applicable                                                                                        Importance to Status                        Finding/Observation                              Disposition Number                SR(s)                                                                                            Application The ANO-1 Fire PRA model ISLOCA logic was revised to capture the potential for a human performance error in the restoration of the DHR isolation valve CV-1400/1401.
LHF1LPITNA was added to the ISLOCA logic to address this issue.
Analyst reviewed Calc. PRA-A1-01-001S08.
There was no reference made to                The value for this pre-initiator was surveillance tests at power, if applicable, in calculated using the following which the ISLOCA pathway configuration        formula.
would be changed from its routine              Unavailability = 4
* TRT / 365*24.
configuration and alignment, e.g., 2 isolation valves instead of 3. Common cause                Where, TRT is time for the valve to This issue was mechanisms were discussed as not being          be in open position per test.      resolved and applicable when they should have been          Per OP-1104.004 Supplement 1, the    therefore has no included.                                                                            impact on STI A1-3931    Resolved  IE-C14                                                  acceptance criteria for opening this valve is 11.8-15.9 seconds.          change evaluations Possible Resolution: Reference testing                                              performed in procedures and their frequency in order to    Assuming that it takes approximately accordance with the more accurately account for the time when      the same time to close the valve, an assumption of 2 minutes per          SFCP.
the ISLOCA pathway is in a different configuration (i.e., 2 valve isolation instead surveillance is a reasonable estimate of 3). Consider common cause failure          for TRT.
mechanisms, which also are related to          Additional references and
                              'state-of-knowledge' correlation (see          discussions added to the ISLOCA QU-A3).                                        documentation (PSA-ANO1                                                                              AS-02) provided to address the CCF and surveillance testing.
State of knowledge correlations are contained in sections 4.3 and Appendix A of the ISLOCA NB that is being developed for Rev. 5.
to 1CAN031801 Page 32 of 102 MCR                Applicable                                                                                      Importance to Status                          Finding/Observation                          Disposition Number                SR(s)                                                                                          Application The capability of secondary system piping appears to be such that once the isolation                                        This issue was valves fail, the low pressure piping                                              resolved and automatically fails.                        Additional discussions have been    therefore has no added to the ISLOCA notebook        impact on STI A1-3932    Resolved  IE-C14    Possible Resolution: State that a            (PSA-ANO1-01-AS-02) to address      change evaluations conservative approach was taken by          the conservatisms.                  performed in assuming automatic failure of secondary                                          accordance with the piping once it is exposed to high pressure,                                      SFCP.
either via leak or rupture.
In reviewing the actual fault tree model, it was seen that the CCF terms were incorporated per the system modeling documentation (Supplement 11) and CCF                                            This issue was documentation (Supplement 4). However,                                            resolved and there was an inconsistency with what was    The CCF event probabilities were    therefore has no stated in Supplement 4 and what was          updated via PSA-ANO1-01-DA-02.      impact on STI A1-3936    Resolved  SY-B4 incorporated in the PRA model.              These updated probabilities are used change evaluations in quantification.                  performed in Possible Resolution: Please resolve                                              accordance with the discrepancy between what was                                                      SFCP.
recommended in the common cause calculation and what was done in the PRA model.
to 1CAN031801 Page 33 of 102 MCR                Applicable                                                                                          Importance to Status                        Finding/Observation                              Disposition Number                SR(s)                                                                                              Application The impact of the state of knowledge correlation on the ISLOCA model was evaluated by developing a simple tree with two tops. The first top is an AND gate of the two check valve failures tied to the same type code. The second top is an AND gate of the two check valves not tied to the same type code. The tree is quantified and the point estimate mean of both tops is essentially the same (as expected). UNCERT is The state of knowledge correlation was not    run for both tops using the same accounted for where it would make a            sample size and the same seed.          This issue was significant difference, i.e., the ISLOCA      The Uncertainty mean for the            resolved and analysis omitted common cause failure of      correlated check valves is              therefore has no check valves. Document PRA-A1              significantly higher than for the non-  impact on STI A1-3946    Resolved  QU-A3    001S08 was reviewed to confirm this.          correlated valves. This test shows      change evaluations Possible Resolution: Consider common          that a correlation factor of 1.42E-6 is performed in cause failure mechanisms, which also may      needed to account for the SOKC          accordance with the be related to 'state-of-knowledge' correlation impact.                                SFCP.
(see QU-A3).                                  Four events are added to the model to address this change:
                                                                                    %FRDH14ASOK Shared Failure of DH-14A and DH-13A due to State of Knowledge Correlation 1.420E-06
                                                                                    %FRDH14BSOK Shared Failure of DH-14B and DH-13B due to State of Knowledge Correlation 1.420E-06
 
Attachment 2 to 1CAN031801 Page 34 of 102 MCR              Applicable                                                                                        Importance to Status                        Finding/Observation                              Disposition Number                SR(s)                                                                                            Application
                                                                                      %FRDH17SOK Shared Failure of DH-14B and DH-17 due to State of Knowledge Correlation 1.420E-06
                                                                                      %FRDH18SOK Shared Failure of DH-14A and DH-18 due to State of Knowledge Correlation 1.420E-06 A1-3946 (continued)                                                                    NUREG/CR-5102, Interfacing Systems LOCA: Pressurized Water Reactors, was used to determine if a common cause failure of check valve leakage or rupture is needed for the ISLOCA. Appendix B of this NUREG evaluates the failure probability of multiple failures and determined that a CCF event is not necessary.
Section 3.0 of the Quantification notebook (PSA-ANO1-01-QU, A review of the Integration and                ANO-1 Integration and Quantification Quantification Work Package (PRA-A1        Work Package) documents the This issue was 001S02) and the FORTE Qualification            Integration and Quantification resolved and Engineering Report (SA-01-001-01) did not      Analysis Method. This section therefore has no reveal any documented software or              includes a table with the name of impact on STI A1-3948    Resolved  QU-F5    quantification limitations that would impact  each software used, what function it change evaluations applications.                                  was used for, and the documented performed in qualification package for that Possible Resolution: Document any known                                              accordance with the software. (FTREX was quant engine quantification limitations, and if none, state                                      SFCP.
used) Section 5.4 of the that there are no known limitations.          Quantification notebook describes limitations of the quantification methodology.
to 1CAN031801 Page 35 of 102 MCR                Applicable                                                                                        Importance to Status                          Finding/Observation                            Disposition Number                  SR(s)                                                                                          Application Attachment E of the Summary Report (PRA-      Section 6 of the Quantification A1-01-001R1) was found to list significant    notebook (PSA-ANO1-01-QU,              This issue was accident sequences and basic events.          ANO-1 Integration and Quantification  resolved and Attachment C lists the top 25 cutsets.        Work Package) documents the            therefore has no QU-F6,                                                definition of significant basic event  impact on STI A1-3949    Resolved            Possible Resolution: Provide definitions for LE-G6                                                (6.3), significant cutset (6.2) and    change evaluations risk significant basic events, cutsets, and  significant accident sequence (6.1). performed in accident sequences within the Summary        The definitions and corresponding      accordance with the Report that comport with those listed in      relevant model results fully satisfy  SFCP.
Section 1-2.2 of the ASME Standard.          the SR QU-F6 requirements.
No documentation of modeling uncertainties is provided in any of the internal flooding notebooks.
Possible Resolution: Identify the applicable This is a modeling uncertainties and document these documentation uncertainties in the appropriate flooding issue and additional IFPP-B3,  notebooks to satisfy the requirements of this discussion of IFEV-B3,  SR.
This is a documentation issue and      flooding sources of IFQU-B3, A1-5886  Unresolved            For example, the flood area uncertainties    additional discussion will be included uncertainty is not IFSN-B3,  could include the potential for leakage      in the IFA documentation.              expected to impact IFSO-B3,  through penetrations that are not sealed as                                          STI evaluations IFQU-A7 designed, undocumented flowpaths                                                    performed in between areas, and equipment that is more                                            accordance with the susceptible to water intrusion than                                                  SFCP.
expected. These uncertainties could be discussed within PSA-ANO1-01-IF-WD and shown to be minimal based on confirmatory walkdowns.
to 1CAN031801 Page 36 of 102 MCR                Applicable                                                                                      Importance to Status                          Finding/Observation                          Disposition Number                  SR(s)                                                                                          Application The calculation of HEPs is documented in the quantification notebook (PSA-ANO1                                IF-QU), attachment A. The HEPs are quantified using the EPRI HRA calculator and generally comply with the Capability Category II post-initiator HEP SRs noted in the HR-E, HR-F, and HR-G HLRs.
However, the following discrepancies are noted:                                                                            Additional HR-E3, HR-E4, and HR-G5: Operator                                              discussion of walkthoughs or talkthroughs of the                                              flooding HEPs, or flooding isolation actions were not                                            adjustments to performed.                              Additional discussion will be included some flooding HFE in the IFA documentation and          timing or execution HR-G4: HFE timing assumptions for A1-5887  Unresolved  IFQU-A5                                              adjustments to some flooding HFE      steps is not groups of scenarios is not consistently timing will be implemented as          expected to impact applied. Some utilize the most limiting needed.                                flooding results, but time and others utilize average time is expected to be available.
assessed in case-HR-G6: A consistency check of the HEPs                                          by-case STI was not performed.                                                              evaluations.
HR-H2: Execution steps for several HFEs include execution recovery steps labeled self-review. These do not appear to be actual procedural steps and there does not appear to be a valid cue that would initiate the recovery. 'self-review' is typically considered in the cognitive portion of an HFE only.
 
Attachment 2 to 1CAN031801 Page 37 of 102 MCR                Applicable                                                                                    Importance to Status                          Finding/Observation                        Disposition Number                  SR(s)                                                                                        Application Possible Resolution: Perform and document talkthroughs of the flooding mitigation actions. Clarify timing assumptions for HFEs and state the basis for whether the most limiting or average A1-5887                          system window will be utilized in the (continued)                      analysis. Determine whether valid cues and procedural steps are present to justify execution recoveries and document the basis for those recoveries. Also perform and document a consistency review of the flooding HEPs.
The lack of a LERF model may impact decisions pertaining to STI changes.
The Rev. 6 model LERF was not considered in the internal                                        update is expected flooding PRA.                                                                  to use the LERF considerations will be included Possible Resolution: Perform an evaluation                                      integrated model to A1-5889    Unresolved IFQU-A10                                              in the IFA analysis and of LERF due to flooding. When performing                                        quantify LERF documentation.
this evaluation, determine if any LERF                                          including internal sequences need to be modified.                                                  flooding. The impact of this finding is expected to be assessed in case-by-case STI evaluations.
to 1CAN031801 Page 38 of 102 MCR                Applicable                                                                                    Importance to Status                          Finding/Observation                          Disposition Number                  SR(s)                                                                                        Application The internal events PRA model, updated by FRANX to include flooding impacts,                                              The incorrect HEP properly considers both flooding impacts                                        is expected to result and random equipment failures and human in a very small errors.                                                                          change to the However, at least one instance was                                              flooding results, as identified in which a modified internal events                                  the majority of HEP was not properly included in the          The impacted HEPs will be updated failures for all flood A1-5890  Unresolved  IFQU-A8  integrated model. See event EFW-HFC-FO-        as needed in the IFA model and    areas were T41BX, which was supposed to be set to        documentation.                    captured.
1.0 in all auxiliary building flooding                                          However, impact of scenarios that occurred at elevation                                            this finding is 386 feet and lower.                                                              expected to be Possible Resolution: Review the process                                          assessed in case-used to incorporate the modified HEPs into                                      by-case STI evaluations.
the internal flooding model and re-quantify using the corrected HEP values.
The internal events PRA model, updated by FRANX to include flooding impacts,                                              The incorrect HEP properly considers both flooding impacts is expected to result and random equipment failures and human                                          in a very small errors.                                                                          change to the Attachment A of the quantification notebook                                      flooding results, as identifies a flooding HEP to be included in                                      the majority of The impacted HEPs will be updated certain flooding scenarios (IFL1-HFC-FO-                                        failures for all flood A1-5891  Unresolved  IFQU-A8                                                  as needed in the IFA model and CTM, to be included with initiating events                                      areas were documentation.
                                %FLAB3350020CTM and                                                              captured. However,
                                %FLAB3350038CTM). However, this HEP                                              impact of this is not included in the flooding PRA model.                                      finding is expected to be assessed in Possible Resolution: Review the process used to include the new flooding HEPs and                                        case-by-case STI ensure that all intended events were                                            evaluations.
included.
to 1CAN031801 Page 39 of 102 MCR                Applicable                                                                                      Importance to Status                          Finding/Observation                          Disposition Number                  SR(s)                                                                                          Application This is a QU-D is not met: There is no evidence that                                        documentation the sequence, cutset, importance, and                                              issue and additional consistency reviews, nonsignificant cutset                                        discussion of review, or the comparisons to other plants  This is a documentation issue and      flooding results is A1-5892  Unresolved  IFQU-A7  were performed for the internal flooding    additional discussion will be included not expected to analysis.                                  in the IFA documentation.              impact STI Possible Resolution: Perform the specific                                          evaluations analyses and evaluations indicated in the                                          performed in QU-D SRs.                                                                          accordance with the SFCP.
This is a documentation issue and additional A parametric uncertainty analysis was not                                          discussion of performed as required by QU-E1.            This is a documentation issue and      flooding results is A1-5893  Unresolved  IFQU-A7                                              additional discussion will be included not expected to Possible Resolution: Perform and            in the IFA documentation.              impact STI document a parametric uncertainty analysis.                                        evaluations performed in accordance with the SFCP.
to 1CAN031801 Page 40 of 102 MCR                Applicable                                                                                  Importance to Status                          Finding/Observation                          Disposition Number                  SR(s)                                                                                      Application The HEP seed values are expected to have It does not appear that HFE values were                                        minimal or no seeded to a higher level to ensure that                                        impact on the cutsets with multiple HFEs are not truncated                                  flooding results, as as required by QU-C1.                        The impacted HFEs will be updated the risk-significant A1-5894  Unresolved  IFQU-A7                                                as needed in the IFA model and    dependencies were Possible Resolution: Perform a systematic    documentation.                    captured.
process to identify all multiple-HFE                                          However, impact of combinations that need to be considered for                                    this finding is the HEP dependency analysis.                                                  expected to be assessed in case-by-case STI evaluations.
to 1CAN031801 Page 41 of 102 MCR                Applicable                                                                                      Importance to Status                          Finding/Observation                            Disposition Number                  SR(s)                                                                                          Application Screening for scenarios was performed within PSA-ANO1-01-IF-AS for scenario groups that had very long operator response times (e.g. greater than 3 hours) due to the relative certainty that operators would be successful in isolating prior to equipment damage. No quantitative basis was provided in this case.
Individual flood initiators were typically not                                    Minimal or no screened out, except for maintenance                                              impact on the related floods. In the case of maintenance                                        results. This is a related floods (see section 2.1.1 of PSA-      This is a documentation issue and  documentation ANO1-01-IF-IE), the screening criteria does    additional discussion of potential  issue only. This IFEV-A8, A1-5895  Unresolved            not meet the numerical threshold of IE-C6. maintenance-related floods flooding Unresolved F&O IFEV-A7 mechanisms will be included in the  does not impact STI Maintenance event screening is currently      IFA documentation.                  evaluations documented as a ~1E-05/yr frequency.                                              performed in Even with the proposed 0.1 factor to                                              accordance with the account for maintenance that breaches the                                          SFCP.
pressure boundary, the criteria are still not met.
Possible Resolution: Perform a more detailed review of maintenance practices and update documentation to show these events meet the screening criteria. Use scenario specific isolation failure data if necessary.
to 1CAN031801 Page 42 of 102 MCR                Applicable                                                                                          Importance to Status                          Finding/Observation                            Disposition Number                  SR(s)                                                                                              Application The discussion of area 34 in PSA-ANO1-01-IF-AS does not appear to include the U2 SW return line in that area. This line needs to be included and the discussion should include the potential for a multi-unit event.
The SW return header is a unit 2 pipe and will require action by U2 operators to realign SW to an alternate discharge path to stop                                            The impact of this the release and to prevent a loss of SW and    Potential initiating events, including finding is expected IFSN-A11,  therefore a unit trip on U2.                  multi-unit scenarios, will be updated A1-5896  Unresolved                                                                                                  to be assessed in IFEV-A4                                                  as needed in the IFA model and Possible Resolution: Revisit potential                                                case-by-case STI documentation.
initiating events in area 34 to determine if                                          evaluations.
the U2 SW return line is applicable. Update accident sequence and initiating events as appropriate. Include operator response for isolation and realignment of SW discharge.
Consider whether a U2 loss of SW event should be included in the U2 flood quantification.
QU-B1: FRANX 4.2 Beta 2 was used for quantification. This software is not                                                  Additional software approved and final tested software. Since                                            testing is not EPRI software testing was not completed,                                              expected to impact this software may contain bugs that would      This is a documentation issue and      the internal flooding A1-5897  Unresolved  IFQU-A7  cause quantification results to be incorrect. additional discussion will be included results or STI Possible Resolution: Utilize approved "final"  in the IFA documentation.              evaluations version of software for quantification or                                            performed in provide a review of the software version                                              accordance with the used that demonstrates that results are                                              SFCP.
acceptable.
to 1CAN031801 Page 43 of 102 MCR                Applicable                                                                                        Importance to Status                          Finding/Observation                          Disposition Number                  SR(s)                                                                                          Application Flood specific post-initiator HFEs were developed to support the internal flood quantification. These HFEs were documented in PSA-ANO1-01-IF-QU This is a Attachment 1. The industry standard HRA                                            documentation calculator was used to perform the analysis.                                        issue and is not HR-G7: Dependency analysis for multiple                                            expected to impact This is a documentation issue and IFQU-A5,  HFEs appears to have been performed.                                                the internal flooding A1-5898  Unresolved                                                        additional discussion will be included IF-QU-A7  However this analysis was not documented.                                          results or STI in the IFA documentation.
Also, it is not clear whether the dependency                                        evaluations analysis includes both IE and IF.                                                  performed in Possible Resolution: Perform and                                                    accordance with the document dependency analysis for multiple                                          SFCP.
HFEs. Ensure the analysis includes combinations of internal events and internal flooding HFEs when applicable.
to 1CAN031801 Page 44 of 102 Table 3 List of SRs Assessed as CC-I in the ANO-1 Internal Events PRA Model SR                    Topic                                            Status                                Importance to Application Updated to CC-II in model revision 5p0. Accident This issue was resolved and therefore sequences consider contributors to large early release.
Accident progression level of                                                                    has no impact in STI change LE-C1                                        Containment challenges for DCH and H2 burn are detail for containment challenges.                                                              evaluations performed in accordance compared with plant specific containment fragility curves with the SFCP.
(See Tables 6.11-2 and 6.11-3 WCAP-16341-P).
Updated to CC-II in model revision 5p0. Generic model extended to include SAMG mitigation and recovery actions. Actions limited to those that may impact LERF.
Actions that impact LERF were determined to be depressurization of the RCS, re-establishment of off-site  This issue was resolved and therefore Treatment of feasible operator power and bumping the pump. ANO-1 guidance to allow      has no impact in STI change LE-C2  actions after onset of core bumping the pump is an important contributor to LERF and    evaluations performed in accordance damage.
its basic event value is studied parametrically. Human      with the SFCP.
action to depressurize the RV following core damage is considered. Recovery of offsite power is considered based on ANO-1 data for LOSP recovery. No additional core damage recovery factor is included.
Conservative treatment is expected to No maintenance or operator actions are credited in this have minimal impact on the results.
evaluation for recovering initially failed High Pressure LE-C3  Credit for repair of equipment.                                                                  However, impact of this finding is Injection (HPI) or any other equipment that would be expected to be assessed in case-by-capable of recovering the core within the reactor vessel.
case STI evaluation.
Updated to CC-II in model revision 5p0. Beneficial failures This issue was resolved and therefore Realistic estimation of significant and outcomes considered for PSV fail to reseat.            has no impact in STI change LE-C4  accident progression sequences Evaluation based on steam cycles only; sensitivity studies  evaluations performed in accordance resulting in large early release.
provided. No fission product scrubbing considered.          with the SFCP.
to 1CAN031801 Page 45 of 102 SR                  Topic                                        Status                                Importance to Application The success criteria for the systems are the same as DCH and hydrogen burn contributors Level 1. Additional success criteria are developed for are not risk significant for most plants.
Containment Isolation. Conservative bounding analyses System success criteria for                                                                  Conservative treatment is expected to used for DCH and hydrogen burn contributors.
LE-C5  significant accident progression                                                            have minimal impact on the results.
Conservative realistic methods are used for PI-TI-SGTR.
sequences.                                                                                  However, impact of this finding is SG tubes considered with an average flaw distribution.
expected to be assessed in case-by-This likely overstates impact since few tubes have actually case STI evaluation.
been plugged.
Effects on operator actions and component availability in harsh The approach taken is WCAP-16341-P, which is an environments are not expected to be extension of NUREG/CR-6595. The NUREG considers significant for LERF for plants with operator action to depressurize without identifying a Credit for continued equipment                                                              large dry containments such as specific analysis to support PORV operation in the post LE-C9  operation or human actions under                                                            ANO-1. Therefore, conservative core damage environment. The approach taken by ANO-1 adverse environments.                                                                        treatment is expected to have minimal does not address PORV operation in the post core impact on the results. However, damage environment but this is consistent with the impact of this finding is expected to NUREG.
be assessed in case-by-case STI evaluation.
Credit for continued equipment operation or human actions under No requirements associated with CCI, credit is precluded LE-C10  adverse environments during                                                                  See above.
by LE-C9.
significant accident progression sequences.
Credit for continued equipment  Updated to CC-II in model revision 5p0. The model          This issue was resolved and therefore operation or human actions that  considers CHR containment failures as LATE unless the      has no impact in STI change LE-C11 could be impacted by            event is preceded by loss of recirculation cooling. In that evaluations performed in accordance containment failure.            event, SI is already inoperable and no credit is taken. with the SFCP.
Credit for continued equipment                                                              Conservative treatment is expected to operation or human actions after                                                            have minimal impact on the results.
Detailed review to reduce LERF contribution not LE-C12  containment failure during                                                                  However, impact of this finding is performed.
significant accident progression                                                            expected to be assessed in case-by-sequences.                                                                                  case STI evaluation.
to 1CAN031801 Page 46 of 102 SR                    Topic                                          Status                              Importance to Application Conservative treatment is expected to All core damage events involving a spontaneous SGTR, have minimal impact on the results.
Treatment of containment bypass      PI-SGTR, or a TI-SGTR event were conservatively LE-C13                                                                                                However, impact of this finding is and credit for scrubbing.            assumed to lead to a large early release. No credit for expected to be assessed in case-by-scrubbing is taken for re-defining LERF events.
case STI evaluation.
Conservative treatment is expected to The approach taken is WCAP-16341-P, which is an have minimal impact on the results.
extension of NUREG/CR-6595. Used ANO-1 fragility LE-D1  Containment ultimate capacity.                                                                However, impact of this finding is curves from NUREG/CR-6475 for DCH and hydrogen burn expected to be assessed in case-by-assessments.
case STI evaluation.
Conservative treatment is expected to Impact of containment seals, The approach taken is WCAP-16341-P, which is an          have minimal impact on the results.
penetrations, hatches, and vent LE-D2                                      extension of NUREG/CR-6595. Generic assessment            However, impact of this finding is piping bellows as potential based on ILRT experience.                                expected to be assessed in case-by-containment challenges.
case STI evaluation.
Conservative treatment is expected to The approach taken is WCAP-16341-P, which is an Definition of containment failure                                                              have minimal impact on the results.
extension of NUREG/CR-6595. Location effects not LE-D3  location that affects classification                                                          However, impact of this finding is considered in determining LERF. All containment failures as a large early release.                                                                      expected to be assessed in case-by-considered large.
case STI evaluation.
Conservative treatment is expected to have minimal impact on the results.
Interfacing system failure          All core damage events involving an ISLOCA event are LE-D4                                                                                                However, impact of this finding is probability.                        conservatively assumed to lead to a large early release.
expected to be assessed in case-by-case STI evaluation.
Secondary side isolation            The analysis has used a conservative evaluation          Conservative treatment is expected to capability for accident progression  (WCAP-16341-P) of secondary side isolation capability for have minimal impact on the results.
LE-D5  sequences caused by SG tube          significant accident progression sequences. For most      However, impact of this finding is failure resulting in a large early  sequences secondary isolation is assumed to fail. ANO-1  expected to be assessed in case-by-release.                            plant specific assessment performed by EPRI.              case STI evaluation.
to 1CAN031801 Page 47 of 102 SR                  Topic                                        Status                              Importance to Application Conservative treatment is expected to The approach taken is WCAP-16341-P, which is an have minimal impact on the results.
Analysis of thermally-induced SG extension of NUREG/CR-6595. Models based on LE-D6                                                                                              However, impact of this finding is tube rupture.                    NUREG-1570 and follow-on EPRI reports including reports expected to be assessed in case-by-with plant specific ANO-1 emphasis.
case STI evaluation.
Negligible LERF contribution expected from containment isolation failure. Therefore, conservative The RBI system is modeled realistically. Purge/venting treatment is expected to have minimal LE-D7  Containment isolation analysis. strategies are not identified, actions are considered low impact on the results. However, probability.
impact of this finding is expected to be assessed in case-by-case STI evaluation.
Relative contribution of contributors to LERF (plant This issue was resolved and therefore damage states, accident Updated to CC-II in model revision 5p0. The significance  has no impact in STI change LE-G3  progression sequences, of accident sequences and phenomena are documented.      evaluations performed in accordance phenomena, containment with the SFCP.
challenges, containment failure modes).
to 1CAN031801 Page 48 of 102 3.3    ANO-1 Fire PRA Model 3.3.1    Plant Changes Not Yet Incorporated Similar to the internal events model, as part of the fire PRA evaluation for each STI change request, sensitivity cases are expected to be explored for areas of uncertainty associated with open items (peer review Findings for ASME/ANS PRA Standard CC II or plant changes) that would impact the results of the STI change evaluation, prior to presenting the results of the risk analysis to the IDP. A review of open MCRs against the ANO-1 fire PRA model showed that there are several ECs related to NFPA 805 modifications which have not yet been implemented.
The internal events and internal flooding models are currently being updated with the NFPA 805 modifications. As stated above, each item is expected to be reviewed and assessed during the specific STI request.
3.3.2    Peer Review Facts and Observations The ANO-1 fire PRA model has undergone several peer reviews, including a full scope and three focused-scope peer reviews. These reviews document the model quality and identify any areas with potential for improvement. The following assessments have been performed and documented for the ANO-1 fire PRA model:
In October 2009, a Westinghouse Owners Group peer review was conducted under LTR-RAM-II-10-003 (Reference 11). There was a total of 54 F&Os, which included 41 findings, 13 suggestions, and no best practices. The conclusion of the review was that the ANO-1 Fire PRA methodologies being used were appropriate and sufficient to satisfy the ASME/ANS PRA Standard RA-Sa-2009. The review team also noted that the staff appeared to be applying methodologies correctly.
In May 2012, a focused-scope peer review was conducted by the Kleinsorg Group to assess supporting requirements FSS-G3, FSS-G4, FSS-G5 and FSS-G6 of the ASME/ANS Combined PRA Standard. This focused-scope peer review is documented in Kleinsorg report 0021-0022-005 (Reference 12). The Kleinsorg Group provided a total of 8 F&Os, of which 6 were findings.
In October 2012, a focused-scope peer review was conducted by Kazarians & Associates (Reference 13), which concentrated on the fire modeling for the SRs of FSS-A, C, D, E and H. The review was focused on the fire modeling parts of the Fire PRA. Only suggestions were provided from this peer review.
In June 2014 a focused-scope peer review was conducted by Curtiss-Wright Scientech (Reference 14) which concentrated on fire HRA development and was performed against RG 1.200, Rev. 2. The peer review found that the ANO Unit 1 fire HRA was performed consistent with the guidance set forth in NUREG/CR-1921, Attachment B, Detailed Quantification of Fire Human Failure Events Using the EPRI Fire HRA Methodology.
There were 5 F&Os as a result of this review, 3 of which were findings.
to 1CAN031801 Page 49 of 102 3.3.3    Consistency with Applicable PRA Standards As discussed in Section 3.1, the ANO-1 Fire PRA model was updated in 2016. Per Entergy procedures, all Entergy PRA models are required to meet current industry standards for PRA model development and documentation. Specifically, the Entergy PRA guidelines were developed to attempt to meet the ASME/ANS PRA standard (Reference 4) CC II of all SRs.
Current Entergy PRA documentation includes an individual self-assessment that documents how each HLR and SR is met.
NUREG/CR-6850 guidance was the primary methodology used for the development of the fire PRA. The updated fire PRA in some cases used methodologies that extend beyond the guidance of NUREG/CR-6850. These methods, used in the ANO-1 Fire PRA Self Approval Model and discussed in Table 4, are considered extensions of the NUREG/CR-6850 methods and are documented via reference to approved NEI 04-02 frequently asked questions (FAQs) or other NUREGs. These references are:
NUREG/CR-6850, Supplement 1, Rev. 0, Fire Probabilistic Risk Assessment Methods Enhancements (EPRI 1019259)
NUREG/CR-7150, Vol 2, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE)
NUREG-1921, Rev. 0, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines -
Final Report FAQ 14-0009, Rev. 1, Treatment of Well-Sealed MCC Electrical Panels Greater than 440V The latest full scope peer review for ANO-1 fire PRA model was conducted in October 2009 using RG 1.200, Rev 2. In addition, focused scope peer reviews were performed in 2012 and 2014 on the FSS, fire modeling, and HRA elements of the fire PRA, resulting in a limited number of additional finding-level F&Os. Since then, a model revision was completed which ensured all the Finding-level F&Os from the peer review were addressed properly. Table 4 provides a listing of the finding-level F&Os related to the fire PRA and the acceptability of the finding-level F&Os in relation to this application. The information provided in Tables 4 and 5 summarizes the information provided in Attachment V of the ANO-1 NFPA 805 License Amendment Request (Reference 17) regarding disposition of the peer review Findings. The column labeled Changes to Modeling Elements has been added to provide information relating to changes to the ANO-1 fire PRA model from that provided in the subject letter. Table 5 lists SRs associated with the fire PRA. These SRs were assessed as CC-I only and the table provides the disposition of CC-I acceptability for this application.
As part of the fire PRA evaluation for each STI change request, sensitivity cases would be expected to be explored for areas of uncertainty associated with open items (peer review Findings for ASME/ANS PRA Standard CC II or plant changes) that would impact the results of the STI change evaluation, prior to presenting the results of the risk analysis to the IDP. At present, there are open items associated with the F&Os; however, the disposition of the F&O addresses the impacts to the fire PRA. Therefore, no sensitivity case is required.
to 1CAN031801 Page 50 of 102 Table 4 List of Finding F&Os against the ANO-1 Fire PRA Model Changes to  Importance MCR            Applicable Status                        Finding/Observation                              Disposition                Modeling          to Number            SR(s)
Elements    Application Fire PRA Peer Review Finding PP-A1-01:                                                            This issue Based on the documentation, it is not clear  This F&O is a duplicate to F&O                      was resolved that all inputs to the evaluation were        CF-A1-02. Supporting Requirement                    and therefore considered in determining the failure        PP-A1 is associated with defining the                has no impact See changes as probability. Input parameters include:        global analysis boundary and does not                in STI change N/A  Resolved  PP-A1                                                                                        depicted for tray/conduit, CPT/no CPT, cable type, and    include failure probabilities.                      evaluations CF-A1-02.
cable configuration. This finding is being                                                        performed in See the response to F&O CF-A1-02 for                accordance assigned against A1 because the lack of the actions to address this comment.                with the documentation reveals that inputs were appropriately used in all cases.                                                                  SFCP.
to 1CAN031801 Page 51 of 102 Changes to  Importance MCR            Applicable Status                        Finding/Observation                                Disposition                    Modeling          to Number            SR(s)
Elements    Application SR PP-B2 references NUREG/CR-6850, Chapter 1 for the acceptable criteria for justifying non-rated fire barriers. NUREG/CR-6850 discusses the use of fire compartments as a well-defined enclosed room, not necessarily with fire barriers. ANO references the FHA as a starting point for plant Fire PRA Peer Review Finding PP-B2-01:        partitioning and all barriers (both rated As discussed in Section 2.2 of Plant          and non-rated are defined in the FHA).
Partitioning and Fire Ignition Frequency,    The Plant Partitioning Task (CALC                            the method used to partition is based upon    E-0016-01, R0) assumes that fire fire zones contained within the Fire          protection features will be effective at                  This issue Hazards Analysis. The barriers as            containing a fire under most conditions.                  was resolved described in the Fire Hazards Analysis are    Fire protection features include fire-                    and therefore both rated and non-rated. Without            rated barriers, non-fire-rated barriers,                  has no impact Original reviewing each individual fire zone          active features such as water curtains,                    in STI change A1-5656 Resolved  PP-B2                                                                                              disposition boundary within the Fire Hazard Analysis      and in some cases spatial separation.                      evaluations remains valid.
(FHA), there is no list of credited barriers  The potential failure of a credited                        performed in that are not rated.                          partitioning feature is addressed in the                  accordance ANO-1 needs to provide a list of barriers    multicompartment analysis (MCA).                          with the credited for fire compartment boundaries      The ANO FHA does not include any                          SFCP.
that are not rated and justify the credit for partitioning features, such as partial boundaries. This may be done through the      height walls, that are discussed in evaluation of adequacy included in the        NUREG/CR-6850 as barriers that multi-compartment analysis.                  should not be credited. Nevertheless, this SR remains open as the current analysis only meets Capability Category I and would result in an identical finding upon re-review. The adequacy of the fire barriers is explicitly reviewed as part of the Multi-Compartment and Hot Gas Layer Analysis calculation (PRA-A1-05-009).
to 1CAN031801 Page 52 of 102 Changes to  Importance MCR              Applicable Status                          Finding/Observation                                  Disposition                    Modeling          to Number              SR(s)
Elements    Application Fire PRA Peer Review Finding PP-B3-01:
Spatial separation is identified as only        This finding was not addressed in the credited for the Turbine Deck, Fire Zones      partitioning effort; therefore, no change to 197-X and 2200-MM, (Report 0247            the partitioning was performed as a result 0006.03, Rev 1, Attachment A, Note 1).          of this finding. The scenario development The note specifies that 1) no significant      and MCA associated with these areas PRA components are located in the vicinity      concluded that the spatial separation (lack of the interface between these fire zones,      of an actual barrier) did not impact fire and 2) the open turbine building and large      results. No fires were judged to credibly associated volume will preclude any            breach the spatial separation and no hot significant fire spread between these          gas layer potential exists.
zones. Per NUREG/CR-6850, Volume 2,            ANO has two areas (four total, two for Pages 1-8, there are several                    each unit) that are separated into Unit 1 considerations for the basis for using          and Unit 2 fire zones that have no fire While this spatial separation as a boundary. The          barrier between the units. The turbine issue remains presence of PRA equipment, although            deck is separated into Zones 197-X for Unit 1 and 2200-MM for Unit 2. Also, the                    unresolved, important in later portions of the analysis, is Fuel Handling areas are separated into                      there is no not relevant with regard to spatial Zones 159-B for Unit 1 and 2151-A for        Original      impact in STI separation as a boundary. The open and A1-5657 Unresolved  PP-B3                                                    Unit 2. The turbine deck area is very large  disposition    change large volume criterion is necessary, but not and a hot gas layer would not develop due    remains valid. evaluations sufficient. Additional criteria that must be to a fire in this area. The MCA performed                  performed in demonstrated include "minimal combustible for ANO turbine deck fire zones screened                    accordance fuel loads," and "free of ignition sources,"
using the NUREG/CR-6850 process.                            with the among others. Therefore, there is In the Fuel Handling Area, the area is SFCP.
insufficient justification documented to support this SR. In addition, two other fire    relatively large and will not create a hot compartments were identified (159-B and        gas layer. The Fuel Handling areas are 2151-A) that credit spatial separation as a    modeled as full room burn-ups in the compartment boundary. The justification        scenario calculation and the MCA screens for use of spatial separation between these    the fire spread to the other area.
compartments was not explicitly identified      This SR remains open as the current in the reports. There is no apparent            analysis only meets Capability Category I systematic review of PRA physical analysis      and would result in an identical finding units to identify when spatial separation is    upon re-review. While the partitioning used or justified.                              element only meets Capability Category I, the scenario and MCA document that this ANO-1 needs to perform a system review limitation has no impact on results.
of the PRA physical analysis unit to identify and justify when spatial separation.
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Elements        Application The disposition to this F&O no Fire PRA Peer Review Finding PP-B5-01:                                                      longer depicts the Documentation does not identify the active                                                  model of record.
fire barriers that are credited in                                                          The information compartment separation. Through                                                            below better discussions, the intent of the evaluation is                                                depicts the current that where active barriers (e.g., fusible-link                                              model of record.
dampers) are rated as part of an overall During the RAI rated wall, the barrier itself is justified as an process, the adequate fire compartment boundary. This          Active fire barriers such as fire dampers                    This issue model was revised justification needs to be included in the        are credited in the FHA for fire zones                        was resolved to address the documentation. Additionally, the                  and are subsequently included in the                          and therefore issue raised by identification and justification needs to be      physical analysis unit (PAU) definitions.                    has no impact this F&O.
provided for active fire barrier components      Failure of these active fire barriers is                      in STI change N/A  Resolved  PP-B5 that are part of non-rated barriers.              included in the MCA by assuming a        There are no        evaluations Evidence could not be found that a                failure probability of 0.0074 based on    active fire        performed in systematic method to identify/justify active      fire door failure. Non-rated barriers are protection          accordance fire barrier components was performed.            addressed with a failure probability of  systems            with the One method to resolve this item is to            1.0 in the MCA.                          supporting the      SFCP.
provide justification for active fire barrier                                              MCA fire barriers components that are included within an                                                      that require an overall rated barrier and identify all active                                              actuation system fire barrier components as part of non-rated                                                that involves barriers and provide justification why                                                      signals from barriers are adequate (i.e., barrier                                                        cables or a configuration is considered during multi-                                                  detection system compartment analysis).                                                                      as part of any PAU boundary at ANO-1.
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Elements        Application The disposition to this F&O no longer depicts the model of record.
The information below better The issues of non-rated barriers, spatial depicts the current separation, and active barriers, are model of record.
discussed above (PP-B2, PP-B3, PP-B5). Walk-downs of all non-NRC or      During the RAI Fire PRA Peer Review Finding PP-C3-01:      Insurance commitment fire barriers        process, the have been performed to document the      model was revised  This issue Documentation is not well developed for basis for credit taken for fire zone      to address the      was resolved the identification and justification of non-boundaries.                              issue raised by    and therefore rated barriers (see F&O PP-B2-01), spatial this F&O. Using    has no impact separation (see F&O PP-B3-01), and active    Quantification of MCA probability has fire protection    in STI change N/A  Resolved  PP-C3    fire barriers (see F&O PP-B5-01). Also,      conservatively used the door failure drawings, if a      evaluations there is no documentation of the walkdown    probability from NUREG/CR-6850 doorway            performed in required in PP-B7 (see F&O PP-B7-01).        Table 11-3 as the boundary failure separated the      accordance ANO-1 should satisfy the resolution for      mechanism for all zones without                              with the adjacent zone F&Os PP-B2-01, PP-B3-01, and PP-B5-01,      openings to adjacent fire zones. For                          SFCP.
from the exposing and document the process and the results. zones with openings to adjacent zones, zone, the door the boundary failure probability was set failure probability to 1.0 and the volume of the combined was utilized along zones was used for assessing the time with type 2 (Fire to HGL formation (PRA-A1-05-009).
Dampers) and Type 3 (Penetration Seals) summed into one probability.
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Elements        Application If no doorway was located on the drawings, the analysis utilized the Type 2 damper barrier probability combined with Type 3 PP-C3                                                                                          (Penetration (continued)                                                                                      Seals) barrier probability. If the drawings were unclear, a Type 1 (Door failure) probability was conservatively added in addition to Type 2 and Type 3.
Fire PRA Peer Review Finding PP-C3-02:
The fire compartments for ANO (Units 1 and 2) are listed in Table 2-2 of the Plant  A note has been added to Table 2-2 of Partitioning and Fire Ignition Frequency    the Plant Partitioning and Fire Ignition development (ERIN Report                    Frequency calculation (CALC-08-E-                            This issue 0247060006.01, Rev. 3, 10/2/09). For        0016-01, which is the Entergy                                was resolved each fire compartment in the table, it would calculation number for the ERIN Report                      and therefore be useful to identify the unit number or if  referenced in the F&O) to indicate that  No change.          has no impact the compartment is a shared compartment      the compartments without unit            Original            in STI change N/A  Resolved  PP-C3                                                  designators can be identified by the (between the two units). Unit number or                                              disposition        evaluations shared designation will facilitate Fire PRA  reference drawing number.                remains valid.      performed in applications, upgrades, and peer review. It  Those with reference drawings starting                      accordance is recognized that the fire zone numbering  with FP-1 are Unit 1 compartments                            with the generally distinguishes between units,      while those with reference drawings                          SFCP.
though not in all cases.                    starting with FP-2 are Unit 2 ANO-1 should clearly identify the unit      compartments.
number or if the compartment is a shared compartment.
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Elements    Application Fire PRA Peer Review Finding ES-C1-01:
The HRA Notebook (Report 0247                            0006.03-U1, Rev. 0) considers instrumentation in terms of providing cues for Operator actions, and determining        The Fire PRA HRA Notebook (PRA-A1-feasibility. The instrumentation credited is  05-007, which is the Entergy report identified in Appendix A associated with the  referenced in the F&O), Attachment A, HRA Basic Event and the related cue.          provides multiple operator cues for                        This issue Several different methods for providing      performing the operator actions. These                      was resolved operator cues are listed (i.e., different    cues show that the operator actions                        and therefore instrumentation sets). Although it is noted  have sufficient diversity so that failure of                has no impact Original if some of the options are not Appendix R    a single instrument or instrument train                    in STI change N/A  Resolved  ES-C1                                                                                              disposition instrumentation, it should be clearly        will not prevent the operators from                        evaluations remains valid.
indicated which option is the credited        performing the action. Attachment B                        performed in instrumentation. See F&O ES-D1-01 and        provides a simulator review of the                          accordance ES-D1-02. Therefore, SR ES-C1 is judged      indicators on the control panels and the                    with the as not met.                                  reliance on instruments. In addition, the                  SFCP.
ANO-1 needs to identify instrumentation      major operator actions are driven by the relevant to operator actions for HFEs to      EOPs. EOPs do not list specific account for the context of fire scenarios in  instruments for performing the actions.
the Fire PRA to meet SR ES-C1. ANO-1 needs to clearly indicate which option is the credited instrumentation associated with HRA Basic Events.
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Elements      Application The disposition to this F&O no longer depicts the model of record.
Attachment B of the ANO-1 FPRA HRA          The information Fire PRA Peer Review Finding ES-C2-01:          Notebook (PRA-A1-05-007) is Fire          below provides PRA Simulator Review - ANO 1. One          the correct        This issue There is no evidence that a systematic                                                                        was resolved review of indications that could result in an  of the specific items addressed during      information the review was to:                          relating to the    and therefore undesirable operator action were identified                                                                    has no impact and dispositioned. A review of control                                                      reference in the Identify critical indicators where fire                    in STI change N/A  Resolved  ES-C2    room instrumentation should be performed                                                    disposition.
damage to sensing devices, cables or                        evaluations to identify possible areas were spurious          other loop components could result in    Revised Calc -    performed in indications of a single instrument could          misleading information that may          PRA-A1-05-015      accordance mislead the operator into performing              cause significant confusion to the      contains          with the undesirable actions is needed to meet              operators and thereby degrade their      Appendix B,        SFCP.
CC-II.                                            effectiveness in the performance of      OPERATOR tasks that are required.                INTERVIEWS AND SIMULATOR EXERCISES with the relevant information.
Fire PRA Peer Review Finding CS-B1-01:
Section 4.4 of the Component and A review of fires that could result in the loss Cable Selection Report (PRA-A1                            This issue of coordination through the loss of control    003), provides documentation that all                          was resolved power (through fire damage to breaker          circuits and electrical distribution buses                    and therefore control cables) was performed. No specific      credited in the fire PRA have been                            has no impact analysis for this was performed. However,                                                  Original analyzed for proper over-current                              in STI change N/A  Resolved  CS-B1    a review of the circuit design indicates that                                              disposition coordination and protection. A                                evaluations this is unlikely to exist based upon circuit                                                remains valid.
description of the processes used is                          performed in design (multiple fuses) and cable routing.      included via reference to Upper Level                          accordance A specific review of this condition should be  Document ULD-0-TOP-12, ANO Unit 1                            with the performed to confirm control and control        and 2 Electrical Protection/                                  SFCP.
power cables do not preclude operation of      Coordination, Rev. 3.
credited equipment.
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Elements    Application Fire PRA Peer Review Finding PRM-A3-01:
This issue The one-top fire PRA model for ANO-1 was      The ANO-1 fault tree model used for was resolved not complete and thus not available for      FPRA application has been completed.
and therefore quantification or review. Determining        Section 14 of the Fire Scenarios Report has no impact dominant contributors and sequence            (PRA-A1-05-004) provides a detailed      Original in STI change N/A  Resolved  PRM-A3    frequencies beyond the scenario level is      description of the ANO-1 databases        disposition evaluations problematic without a one-top fire model. utilized to generate, document and        remains valid.
performed in Create the one top fire model and            quantify the fire PRA model. This accordance benchmark results against the FRANC          section includes reference to the with the results for use in quantifications and        specific fault tree model used.
SFCP.
review.
Fire PRA Peer Review Finding PRM-B2-01:
The ANO-1 Internal Events Peer The PRA Peer Review for the ANO-1            Review was performed in August 2009.
internal events PRA was performed the first  The ANO-1 Fire PRA Peer Review was week of August 2009. As such, there has      performed in late October 2009. Based been insufficient time to reconcile the F&O  on the limited time between the peer                    This issue that could have an impact on the Fire PRA. reviews, ANO did not have time to                        was resolved However, the ANO PRA team has                incorporate internal events F&Os before                  and therefore reviewed all of the F&Os from the ANO-1      the Fire PRA review. As stated in this                  has no impact internal events PRA peer review. They        F&O, a limited number of internal        Original in STI change N/A  Resolved  PRM-B2    have identified five F&Os that have the      events findings were determined to        disposition evaluations potential to impact the fire PRA, and        impact the Fire PRA. These F&Os          remains valid.
performed in developed action plans for their disposition. were subsequently incorporated in the                    accordance The F&Os from the internal events ANO        FPRA.                                                    with the PRA peer review that could impact the Fire                                                            SFCP.
Table 2 provides details about the PRA have not yet been implemented.
internal events F&Os. The details ANO-1 needs to implement the action plan provided in Table 2 include the status of that has been developed to reconcile the each F&O and each findings potential ANO-1 internal events PRA F&Os that impact.
could impact the fire PRA.
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Elements    Application Fire PRA Peer Review Finding PRM-B5-01:  Section 4.5 of the ANO-1 FPRA There is no evidence of a review the      Component and Cable Selection Report accident sequence models to determine    (PRA-A1-05-003) discusses the Plant                    This issue whether sequences need to be added or    Response Model, including a dedicated                  was resolved changed. This SR requires a REVIEW of    section discussing success criteria.                  and therefore the corresponding accident sequences and  Appendix D discusses various accident                  has no impact Original there is no objective evidence that this  sequence types and how they would or                  in STI change N/A  Resolved  PRM-B5                                                                                      disposition review was performed, although there does would not apply for the FPRA.                          evaluations remains valid.
not appear to be any modified accident    Additional comments on the internal                    performed in sequences and the staff confirmed that no events model are discussed in                          accordance sequences were modified.                  Appendix F. The details provided in the                with the ANO-1 needs to document the review of    component and cable selection                          SFCP.
the corresponding accident sequences for  document satisfy the PRM-B5 addition or modification.                Supporting Requirement.
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Elements        Application The disposition to this F&O no longer depicts the Traditional multi-point heat release rate  model of record.
treatment was not applied in the ANO-1    During the RAI analysis. Rather than the repetitive      process, the Fire PRA Peer Review Finding FSS-C1-01:                                                  model was revised analysis inherent in a multi-point heat Assignment of characteristics to fire        release rate treatment, the Conditional    to address the scenarios does not meet CC II. In order to    Probability of Propagating Fire factors    issue raised by meet CC II, a two-point fire intensity model  specified for vented panels provide a      this F&O.
must be used to assign characteristics to    multi-point treatment for vented panels    The following the ignition source. Furthermore, per        based on a split fraction developed from                        This issue information was Note 2 of Table 4-2.6-4 (c) of the            the EPRI Fire Events Database. This                            was resolved obtained from the ASME/ANS RA-Sa-2009 Standard, CC II          split fraction specifies the fraction of                        and therefore ANO-1 SE requires, as a minimum that the              fires impacting only the ignition source                        has no impact (ML16223A481).
determination of minimum fire intensity      panel versus those fires which impact                          in STI change N/A  Resolved  FSS-C1 capable of causing fire spread and/or        targets within the zone of influence of    In PRA RAI 01.c      evaluations damage to at least one member of the          the panels. This approach provides a      (Reference 21),      performed in target set. Then the two-point fire intensity definitive means of differentiating        the NRC staff        accordance model is applied to characterize the          between significant and limited fires that requested that the  with the damaging fires (i.e., fires above the        will not be significantly impacted by      licensee explain    SFCP.
minimum damage intensity). Therefore,        potential future refinements in ignition  its approach and this SR is judged to be met at CC I.          frequency and heat release rate.          whether the split A two-point fire intensity model must be                                                fraction referred to Section 16 of the Fire Scenarios Report    in the resolution to used to assign characteristics to the (PRA-A1-05-004) discusses the use of      FSS-C1-01 ignition source to meet CC II.
generic fire modeling versus detailed      implied that the fire modeling and justifies the ANO-1      "Panel Factors approach for the FPRA application.        method" (Reference 105),
not accepted by NRC was used.
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Elements        Application In its response to PRA RAI 01.c (Reference 10),
the licensee clarified that that the "Panel Factors" method was not used but rather generic fire modeling treatments (GFMTs) were used to define a three-point fire model to meet SR FSS-C1. The licensee explained that the first fire in FSS-C1                                    the three-point fire (continued)                                treatment is a non-severe fire in which the source panel and the cables terminating at the source panel are damaged but the nearest target is not damaged.
The second fire in the three-point fire treatment is a severe fire in which all targets within the 98th percentile zone of influence (ZOI) are impacted.
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Elements        Application The third fire of the three-point fire treatment is a fire that results in a hot gas layer (HGL) exceeding a 80 degree Centigrade
(&deg;C) criterion in which all targets in the fire zone are conservatively assumed to be damaged. These three fire models are discussed in the licensee's response to FM RAI 01.f discussed FSS-C1 in SE Section (continued) 3.4.2.3. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that it used a multiple-point fire intensity and duration model encompassing low likelihood but risk-contributing fire events consistent with the requirements in the PRA standard.
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Elements      Application Fire PRA Peer Review Finding FSS-C2-01:
The generalized models used to support most of the significant fire scenario evaluations use peak heat release rates.
Since the ANO-1 FPRA Peer Review, For example, 8 of the top 10 fire scenarios ANO has made several refinements listed in the CDF quantification results within the reviewed methodology to presented in Appendix A of the Summary remove some conservatism and reduce Report (Entergy Report 0247060006.06-overall CDF and LERF. These U1, Rev. 0, 9/11/09) are for "Base refinements include:
Scenarios," which are the equivalent of a fire safe shutdown analysis exposure fire in        developing more detailed fire which everything in the compartment is              scenarios                                          This issue assumed to fail at the compartment                                                                      was resolved refining the components failed  The disposition to frequency without considering time-                                                                      and therefore within the scenario              this F&O no dependent fire growth. Time-dependent                                                                    has no impact longer depicts the N/A  Resolved  FSS-C2 fire growth is considered on a limited basis,      refining the fire HRA events and model of record.
in STI change such as for the main control room                    JHEPs.                                              evaluations See revised abandonment scenario and for ventilated                                                                  performed in The use of fire growth curves are not  disposition for cabinets that are located in zones equipped                                                              accordance part of the Generic Fire Modeling      F&O FSS-C1-01.
with automatic detection, where credit may                                                              with the Treatments used for ANO-1. Section 16 be taken for suppression by the fire brigade                                                            SFCP.
of the Fire Scenarios Report (PRA-A1-prior to sustaining external target damage 05-004 which is Entergy calculation for (Fire Scenario report, Entergy Report 0247-ERIN Report 0247-06-0006.05-U1) 06-0006.05-U1, Rev. 0, 9/11/09). Therefore discusses the use of generic fire this SR is met at CC I.
modeling versus detailed fire modeling, Time-dependent fire growth should be          and justifies the ANO approach for the considered for more of the significant fire  FPRA application (and only meeting scenarios, which are mostly evaluated        Capability Category I).
using peak heat release rates to meet CC II. Expand use of time-dependent fire growth to additional significant fire scenarios as appropriate.
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Elements      Application The disposition of this F&O is not accurate in relation to the current model of record.
Multiple zones credit suppression Fire PRA Peer Review Finding FSS-C7-01:
and detection.
A multiple suppression path is modeled for Section 9.0 of the Fire Scenario Report  Section 9.0 of the This issue the cable spreading room. Proper          (PRA-A1-05-004) documents the            Fire Scenario      was resolved modeling appears to have been performed    Credit for Suppression and Detection    Report (PRA-A1-    and therefore for the cable spreading room in the self-  Systems and explicitly outlines how the 05-004)            has no impact assessment for FSS-C7. This information    NSP is calculated for the Cable          documents the      in STI change N/A  Resolved  FSS-C7 should be formally documented with        Spreading Room (including appropriate    Credit for        evaluations appropriate references. Documentation of  references). Explicit credit for        Suppression and    performed in calculation needs to be included in Fire  suppression and detection systems is    Detection          accordance PRA documentation.                        taken for the Cable Spreading Room      Systems.          with the Document the calculation and include      fire scenario only.                      Additionally, the  SFCP.
appropriate references.                                                            MCA/HGL analysis (PRA-A1-05-009) provides information relating to the determination of suppression and detection when credited.
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Elements      Application Fire PRA Peer Review Finding FSS-D1-01:
Simplified fire modeling is performed as described in Attachment B of the ANO -
Unit 1 Fire Scenario Report (Report 0247-Section 16 of PRA-A1-05-004, which is 06-0006.05-U1 Revision 0). There has                                                                  This issue the referenced report in the F&O, been no further area-specific modeling                                                                was resolved explains and justifies the ANO use of  The disposition to done with more sophisticated tools to                                                                  and therefore the Generic Fire Modeling approach      this F&O no determine if significant risk contributors                                                            has no impact instead of a more detailed approach. longer depicts the could be reduced or if the conservative                                                                in STI change N/A  Resolved  FSS-D1                                                This approach is based the Zone of      model of record.
values are bounding for all dominant                                                                  evaluations Influence (ZOI) dimensions for each    See revised scenarios. The potential may exist for                                                                performed in heat release rate bin on the value that disposition for nonconservative scenarios as well as risk                                                              accordance produced the largest distance. In the  F&O FSS-C1-01 reductions in the significant scenarios.                                                              with the absence of specific data, this is a Investigate further into whether or not use                                                            SFCP.
conservative approach.
of more sophisticated modeling tools would change the results for the dominant fire scenarios in the higher risk areas, such as Fire Zone 99-M.
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Elements      Application The disposition to this F&O no longer depicts the model of record.
See revised disposition for Fire PRA Peer Review Finding FSS-D3-01:                                              F&O FSS-C1-01.
Fire growth and propagation within cable                                            Additionally, from trays are not explicitly treated in the HGL                                          the ANO-1 SE calculations. Page 34 of ANO-1 Fire PRA                                              (ML16223A481)
Summary Report says: "This approach                                                  In its response to incorporates conservatisms in the time to                                            FM RAI 01.c(i)      This issue HGL impact that, along with the                                                      (Reference 11),    was resolved conservatism of the heat release rates      Section 16 of the Fire Scenarios Report the licensee        and therefore used, will envelope impacts due to          (PRA-A1-05-004) discusses the use of    stated that it used has no impact additional heat release rates introduced by  generic fire modeling versus detailed  the FLASH-CAT      in STI change N/A  Resolved  FSS-D3    the ignition of cable trays external to the  fire modeling and justifies the ANO    model to calculate  evaluations initial ignition source." Fire growth and    approach for the FPRA application.      the HRR increase    performed in propagation within the cable trays can add  This approach ensures ANO meets only    due to fire        accordance energy to the fire that would affect the HGL the Capability Category I for FSS-D3. propagation in      with the calculations. Although conservatisms in                                              cable trays, and    SFCP.
the Heat Release Rate (HRR) may                                                      that it determined envelope this additional added heat, this is                                        the expanded only assumed.                                                                        vertical and Quantitatively evaluate fire growth and                                              horizontal ZOI propagation within cable trays for the HGL                                          based on the total calculations.                                                                        HRR of the ignition source and secondary combustibles using the methods described in the GMFTs.
to 1CAN031801 Page 67 of 102 Changes to      Importance MCR            Applicable Status                          Finding/Observation                              Disposition                    Modeling              to Number              SR(s)
Elements        Application The NRC staff concludes that the licensee's response to the RAI is acceptable because the FSS-D3                                                                                            licensee properly (continued)                                                                                          accounted for the HRR contribution of secondary combustibles (cable trays) in determining the ZOI.
Fire PRA Peer Review Finding FSS-D7-01:      Section 9.0 of the ANO-1 Fire Scenario    Due to revisions in The assessment of unavailability is, in part, Report (PRA-A1-05-004) details credit      the model, the based on fire protection program controls    for detection and suppression systems. disposition to this for implementing compensatory actions for    Explicit credit for suppression and        F&O no longer out of service systems. The types of          detection systems is taken for the Cable  depicts the model compensatory measures for out of service      Spreading Room (Fire Zone 97-R) fire      of record. The detection and suppression systems are        scenario only. Per Technical              following response  This issue given in the TRM. For detection systems      Requirement for Operation (TRO)            better depicts the  was resolved (Section 3.3.6 of the TRM), hourly roving    3.3.6.B, the failure of a detector in this current Fire PRA    and therefore fire watches are used when less than 50%      zone would require the automatic          model.              has no impact of the detectors in a zone are operable.      suppression system in this area to be                          in STI change N/A  Resolved  FSS-D7                                                                                            Section 9.0 of the For sprinkler systems (Section 3.7.9 of the  declared inoperable per TRO 3.7.9.A,                          evaluations ANO-1 Fire TRM), an hourly fire watch is established if  which requires a continuous fire watch                        performed in Scenario Report detection is operable in the area,            to be established within 1 hour. A                            accordance (PRA-A1-05-004) otherwise, a continuous fire watch is        review of plant maintenance history                            with the details credit for established. Per NUREG/CR-6850,              shows that limited unplanned                                  SFCP.
detection and Appendix P, only continuous fire watches      maintenance has been performed on suppression can be used for crediting availability of    this detector in the past 20 years.
systems. There detection and suppression systems.            Therefore, the unavailability of this      are several fire Consider removing crediting of hourly fire    system is very low and is considered to    zones in which watches and include component-specific        be enveloped by the system unreliability  detection is unavailability data.                          data taken from NUREG/CR-6850.            credited.
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Elements        Application Per Technical Requirement for Operation (TRO) 3.3.6.B, the failure of a detector in this zone would require the automatic suppression system in this area to be FSS-D7                        There is no credit taken for hourly fire declared (continued)                    watch.
inoperable per TRO 3.7.9.A, which requires a continuous fire watch to be established within 1 hour. There is no credit taken for hourly fire watches in these zones.
to 1CAN031801 Page 69 of 102 Changes to        Importance MCR            Applicable Status                          Finding/Observation                            Disposition                    Modeling                to Number            SR(s)
Elements          Application The original F&O disposition does not reflect the current model of record. The following response Fire PRA Peer Review Finding FSS-G2-01:
better depicts the Embedded in the analysis is the                                                        current Fire PRA assumption that a fire is always controlled                                            model.
within 20 minutes such that a hot gas layer (HGL) will not form beyond 20 minutes.                                                The manual non-This assumption is based on the concept                                                suppression that the fire brigade would arrive within                                              probability (NSPms) 20 minutes and successful mitigate the                                                is calculated using effects of the fire within that time by                                                a convolution process which        This issue opening doors or suppressing the fire. This The updated Multi-Compartment/Hot        applies the non-      was resolved is treated in the model as a 1.0 probability.
Gas Layer Analysis (PRA-A1-05-009)      suppression curve    and therefore Issues associated with this assumption uses a distributed manual suppression    to each bin of the    has no impact include: 1) no evidence is provided that the probability based on 20-, 30-, 60-minute applicable            in STI change N/A  Resolved  FSS-G2    HGL temperature would not continue to HGL growth rates. The updated            NUREG/CR-6850,        evaluations increase following the opening of a door, MCA/HGL report develops                  Appendix E (Tables    performed in and 2) this evaluation does not consider non-suppression probabilities based on  E-2 through E-9)      accordance the probability distribution of brigade FAQ 08-0050 (FAQ 50) guidance.          heat release rates    with the response time coupled with the actions that (HRR), with the      SFCP.
would be taken such as what is considered in NUREG/CR-6850 and Frequently Asked                                                  time to hot gas Question (FAQ) 50. Non-conservative                                                    layer varying based screening methodology that may screen                                                  on the heat release significant compartments.                                                              rate associated with each bin. This One possible resolution would be to credit NSPms is multiplied a distributed manual suppression by the fraction of probability based upon actual time for hot the total probability gas layer development.
distribution (severity factor) applicable to each discrete bin of the HRR distribution and summed.
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Elements      Application This final value gives a manual non-suppression probability (NSPms) for each evaluated ignition source. The time to HGL is determined for each heat release rate bin using a curve fit of the available time to hot gas layer data for the various FSS-G2                                    configuration heat (continued)                                release rate (single cable bundle panel, multiple cable bundle panel, motor).
Additionally, the 20-30-60 min timing intervals have been replaced with detailed timing analysis for each fire scenario event tree.
to 1CAN031801 Page 71 of 102 Changes to  Importance MCR            Applicable Status                          Finding/Observation                                Disposition                  Modeling          to Number            SR(s)
Elements    Application Fire PRA Peer Review Finding FSS-G2-02:
The multi-compartment analysis assumes that for each fire scenario, the only source of heat contributing to the hot gas layer is    The updated Multi-Compartment/Hot the heat from the cabinet (based upon a        Gas Layer Analysis (PRA-A1-05-009) 98% HRR). This does not account for            uses more detailed methods for analyzing hot gas layer development                      This issue additional heat due to potential fire spread and growth. The updated methods                          was resolved to other combustibles including cable. It is account for both fire spreading and                      and therefore noted that a conservative (98%) HRR rate additional combustibles.                                has no impact is used for the cabinet. However, it is not                                              Original in STI change N/A  Resolved  FSS-G2    demonstrated that this is bounding (i.e., 1.0  The listed details are in Attachment A of disposition evaluations probability) when considering the HRR over      the Multi-Compartment and Hot Gas        remains valid.
performed in time due to fire spread. This could result in  Layer Analysis. The supplemental                        accordance non-conservative screening of                  information relates to additional hot gas                with the compartments - this may be compounded          layer tables generated for transient                    SFCP.
by the issue identified by F&O FSS-G2-01.      fires, specific steady and peak heat One method to resolve this is to consider      release rate values, and scenarios that the HRR based on generic bounding or            involve secondary combustibles.
compartment specific configuration of cabinets vs. cables and described in NUREG/CR-6850 and FAQ 49.
Fire PRA Peer Review Finding FSS-G2-03:
The analysis assumes (with the exception of the control room) that all barriers have no openings. Cases were identified where other                                                              This issue openings in fire compartment barriers exist,    The updated Multi-Compartment and                        was resolved e.g., 73W. A systematic effort is needed to    Hot Gas Layer Analysis (PRA-A1                      and therefore identify these openings so that they can be    009) methodology addresses openings                      has no impact accounted for in the multi-compartment          between fire compartments.                Original in STI change N/A  Resolved  FSS-G2    analysis. This is a follow-on to the issue                                                disposition identified in F&O PPB3-01. The review          Openings between fire compartments                      evaluations remains valid.
indicates that it is likely on few compartments are identified and their impact evaluated                performed in have such openings.                            for every fixed ignition source analyzed                accordance in the Multi-Compartment Analysis.                      with the Identify these compartments as part of Plant SFCP.
Boundary and Partitioning. Include the openings and impact of openings in the Multi-compartment analysis to justify low significance.
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Elements        Application The disposition to this F&O does not Fire PRA Peer Review Finding FSS-G2-04:                                                reflect the current The hot gas layer analysis is based on an                                              model of record.
actual fire compartment volume (area of                                                The information room times height) where the fire modeling                                              below is provided used to account for the amount of heat                                                  from the            This issue The Multi-Compartment and Hot Gas          MCA/HGL report.
necessary to cause a hot gas layer is                                                                      was resolved Layer Analysis has updated the based upon the available hot gas layer                                                  Electrical Panel    and therefore methodology in the following manner.
volume. To evaluate the hot gas layer, it                                              Fires are assumed  has no impact The new approach assumes the ignition was assumed that the fire was at the floor                                              to be seven feet    in STI change N/A  Resolved  FSS-G2                                                source/fire has a base 8 ft off the floor.
level. This may result in non-conservative                                              high with fires at  evaluations The new methods also calculate room estimates for the amount of heat necessary                                              the top of the      performed in volume above the source for HGL to result in a hot gas layer. This                                                      panel, reducing    accordance impact (Attachment A of PRA-A1                            methodology results in non-conservative                                                the total height of with the 009).
heat requirements to cause a hot gas layer.                                            the room            SFCP.
Adjust the assumed room volumes in the                                                  associated with screening process based upon the                                                        the volume of air available hot gas layer volume as opposed                                              above the panel to the full room volume.                                                                available for hot gas layer heat up by seven feet.
Fire PRA Peer Review Finding FSS-H4-01:
There is no apparent documentation that Attachment D of the Fire Scenarios the technical bases for input values used in Report (PRA-A1-05-004) is a Walkdown                          This issue the fire modeling were validated by plant Workbook. This attachment provides                            was resolved walkdowns or other methods. This SR the basis for FRANC inputs.                                    and therefore requires documentation of a technical basis Attachment A-2 of the Scenarios Report                        has no impact to be established for fire modeling tool                                                Original is the Scenario Development Walkdown                          in STI change N/A  Resolved  FSS-H4    input values given the context of the fire                                              disposition Summary. These two attachments                                evaluations scenarios being analyzed. This was                                                      remains valid.
(with some additional information in                          performed in reported to be performed as part of the other Scenario Report attachments)                            accordance walkdowns for scenario development, but contain the details to support that all                        with the not documented.
scenario development inputs were                              SFCP.
Document that the technical basis for fire  validated by walkdowns.
modeling input values were validated in the context of the fire scenarios analyzed.
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Elements    Application Fire PRA Peer Review Finding IGN-A5-01:
Table 3-2: some IEFs are not calculated on a "per reactor-year" basis. All NUREG/CR-This issue 6850 ignition frequencies should be was resolved updated with reactor critical years in order Fire PRA Plant Partitioning and Fire                    and therefore to obtain the correct ignition frequencies Ignition Frequency Development                          has no impact per this SR. IGN-A5 requires generic fire                                            Original calculation (CALC-08-E-0016-01                          in STI change N/A  Resolved  IGN-A5    ignition frequencies or plant-specific fire                                          disposition Table 3-2) has been changed to show                    evaluations frequency updates on a reactor-year basis.                                            remains valid.
that all bins were updated on a reactor-                performed in This is not done for the following ignition year basis.                                            accordance frequency bins: 1, 4, 8-10, 12-16c, 17-19, with the 21, 23, 26, and 30.
SFCP.
Update the "all-mode" Ignition Frequencies from NUREG/CR-6850 with critical years as opposed to calendar years.
Fire PRA Peer Review Finding IGN-B5-01:
Section 1.1 of Report 0247060006.01 Revision 3 contains only one assumption.
However, a text search of the document resulted in a number of additional instances of assumptions buried in the text. One was                                                          This issue an assumption that the ignition frequencies                                                          was resolved were log-normally distributed, one was the                                                          and therefore The ANO Fire Probabilistic Risk assumption that the compartments were                                                                has no impact Assessment - Plant Partitioning and      Original assigned in accordance with the generic                                                              in STI change N/A  Resolved  IGN-B5                                                Fire Ignition Frequency Development      disposition sources, one was the assumption that                                                                evaluations (CALC-08-E-0016-01) contains an          remains valid.
junction boxes were uniformly distributed,                                                          performed in updated section on assumptions.
and a general assumption for a number of                                                            accordance the events that they occurred at power.                                                              with the All assumptions pertaining to the ignition                                                          SFCP.
frequency calculation should be explicitly captured in Section 1.1, with the possible exception of the "at-power" assumption for individual events. The assumptions should also be reviewed for completeness.
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Elements        Application The disposition to this F&O does not reflect the current model of record.
The information Fire PRA Peer Review Finding CF-A1-01:
below is depicts The tables in NUREG/CR-6850 were used                                                      the current to determine cable failure probabilities.                                                  modeling and use The most conservative value from this table                                                of a new was chosen. The analysis does not              Section 12.0 of the Fire Scenario          methodology.
account for the potential for both inter and    Report (PRA-A1-05-004) discusses The NUREG/CR-        This issue intra cable short probabilities. For            cable failure probabilities. The section 6850 data            was resolved example, item 32-K, LMV101414K was              outlines the application of inter and intra associated with      and therefore assigned a failure probability of 0.3 for intra cable short probabilities.
Hot Shorts has      has no impact cable hot short. This value does not For a circuit with a CPT, a bounding hot    been revised to      in STI change N/A  Resolved  CF-A1    include the potential for inter cable hot short probability of 0.33 is used which    reflect the          evaluations short of .03 (total probability of 0.33).
includes both intra- and inter-cable hot    probabilities        performed in Since the highest failure probability in the shorts. For a non-CPT circuit, a            associated with      accordance table is used - this is typically the intra bounding hot short probability of 0.66 is  NUREG/CR-7150,      with the cable failure probability, the impact of used which includes both intra- and        Joint Assessment    SFCP.
excluding the inter cable failure probability is relatively small.                            inter-cable hot shorts.                    of Cable Damage and Quantification Review cables where failure probabilities of Effects from other than 1.0 are credited and ensure the Fire (JACQUE-appropriate inter and intra cable short FIRE). The use probabilities are applied.
of NUREG/CR-7150 will not affect the original disposition of this F&O.
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Elements        Application The NUREG/CR-6850 tables used for hot shorts are Tables 10-1 and 10-2, which are associated with thermoset      The NUREG/CR-cables. The remaining tables are          6850 data Fire PRA Peer Review Finding CF-A1-02:                                                associated with associated with thermoplastic cables Based on the documentation, it is not clear (Tables 10-3 and 10-4) or armored        Hot Shorts has that all inputs to the evaluation were      cables (Table 10-5). Calculation CALC-    been revised to This issue considered in determining the failure      ANOC-FP-09-00019 identified only 10      reflect the was resolved probability. Input parameters include:      thermoplastic cables for ANO with only    probabilities and therefore tray/conduit, CPT/no CPT, cable type, and  two of these cables associated with      associated with has no impact cable configuration. This finding is being  ANO-1.                                    NUREG/CR-7150, in STI change N/A  Resolved  CF-A1    assigned against A1 because the lack of                                              Joint Assessment Details about parameters pertaining to                        evaluations documentation reveals that inputs were                                                of Cable Damage cable & circuit failure probabilities are                      performed in appropriately used in all cases.                                                      and Quantification documented in the Fire Scenario Report                        accordance Document the specific configuration inputs                                            of Effects from (PRA-A1-05-004). The cable at ANO is                          with the used in justifying the chosen failure                                                Fire (JACQUE-type IEEE-383 and the damage                                  SFCP.
probabilities from NUREG/CR-6850 failure                                              FIRE). The use of threshold for this type is specified in  NUREG/CR-7150 table probability and validate the chosen NUREG/CR-6850 (Section 6 of the          will not affect the probabilities.
Scenario Report). Section 12.0 of the    original disposition Fire Scenario Report addresses the        of this F&O.
failure data applied for CPT/non-CPT circuits.
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Elements    Application Fire PRA Peer Review Finding HRA-A2-01:
To address an excess spray MSO, ANO-1 an HFE, RHF1RCPSXP, was added directly to the model logic. In this logic, RHF1RCPSXP feeds into an AND gate and has a value of "0.0" which in effects kills                                                          This issue The modeling error has been corrected.
the entire logic structure. This HFE shows                                                            was resolved The new event (RHF1RCPSXP) has a up in the TAGBE file tagged as "N2" which                                                            and therefore default value of 1.0. It is only changed means it is ignored. It does not show up                                                              has no impact if changed in the altered events table (in Original anywhere in the HRA report, the                                                                      in STI change N/A  Resolved  HRA-A2                                                FRANC) if it is relevant to a specific    disposition ExcludedEvents file or the AlteredEvents                                                              evaluations case. With the default value of 1.0, it is remains valid.
table so there is no definition or                                                                    performed in does not disrupt quantification of the characterization that is traceable.                                                                  accordance other logic in the AND gate for other Undocumented HFE that appears to be                                                                  with the cases.
impacting logic. If RHF1RCPSXP is not                                                                SFCP.
used, rather than setting its value to 0.0, remove it from the model. If it is a valid HFE, it needs to be identified and fully characterized in the HRA report and the correct value needs to be calculated.
Fire PRA Peer Review Finding HRA-A3-01:    Attachment C of the ANO-1 FPRA There is limited direct evidence that a    Human Failure Events Notebook (PRA-systematic review of fire scenarios was    A1-05-008) contains the systematic performed to identify undesirable operator  review of fire PRA credited operator actions. The Attachment contains the                      This issue action that could result from spurious results of interviews with experienced                    was resolved indications (See F&O ES-C2-01). The ANO-1 operations personnel.                              and therefore evidence in Attachment E of the HRA has no impact report is implicit. A review of each fire  Experienced operators were asked a        Original in STI change N/A  Resolved  HRA-A3    scenarios is needed to identify undesirable series of questions about each HEP        disposition evaluations operator action that could result from      credited in the model. The questions      remains valid.
performed in spurious indications of a single instrument included - description of the action,                    accordance for CC II.                                  which procedures apply, what                              with the Perform a systematic review of fire        instruments/signals are available, time                  SFCP.
scenarios to identify undesirable operator  available and/or required to take action, action that could result from spurious      location (inside or outside control room),
indications of a single instrument, per SR  and if there are any special ES-C2.                                      considerations.
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Elements        Application The FPRA HRA Notebook (PRA-A1                                                                            007), Attachment A, provides multiple operator cues for performing the operator actions. These cues show that HRA-A3                                                  the operator actions have sufficient (continued)                                              diversity so that failure of a single instrument or instrument train will not prevent the operators from performing the action and reduces the likelihood of inadvertent/undesirable actions.
Fire PRA Peer Review Finding HRA-B3-01:                                                  The F&O does not The new human failure events (HFEs) as                                                  reflect the current identified in the FRANC                                                                  model of record.
ALTEREDEVENTS table are not                                                            The information developed as per HRA-B3 (nor identified in                                              below is depicts HRA-B2). These HFEs need to be                                                          the current processed to (1) determine the viability of The HRA events have been removed modeling and use the HFE - can it be performed, are there      from the Altered Events table, except of a new cues, etc., and (2) satisfied the HR          for actions that are set to TRUE in a fire methodology.
supporting requirements (SRs) from            scenario where fire prohibits the                              This issue Section 2 of the ASME/ANS PRA Standard        operator action. The revised FPRA          The ANO-1 HRA      was resolved (internal events). Note that the HFEs        HRA analysis (PRA-A1-05-008 - ANO-1        methodology was    and therefore identified in HRA-B1 (from the internal      Fire PRA Human Failure Events)            revised consistent  has no impact events PRA) would have been assessed at      provides detailed HRA evaluations for      with the approach  in STI change N/A  Resolved  HRA-B3                                                  most of the fire-specific operator        used to address CC III. The table in Appendix A of ANO-1                                                                    evaluations Fire Probabilistic Risk Assessment Human      actions.                                  ANO-2 NFPA 805      performed in Reliability Analysis (HRA) Notebook                                                      LAR RAIs. The      accordance Two of the events are left at screening (Report 0247060006.03-U1, Revision 0),                                                  revised            with the values (QHFSGDEPRES = 0.1 and September 2009 deals with timing and                                                    methodology uses    SFCP.
RHF1ESASRG = 1.0). All events used availability of cues. Since the SR HRA-B1                                                the NUREG-1921 in the FPRA have been developed and HFEs are from the internal events PRA, the                                              methodology with documented using the same methods specific procedure guidance and task                                                    detailed Human used for the internal events HEPs.
analysis are contained in the hfe_cr.xls and                                            Error Probabilities hfe_cp.xls Excel spreadsheets from the                                                  (HEPs) developed internal events ANO-1 PRA. The HR SRs                                                    for each Human from Section 2 of the ASME/ANS PRA                                                      Failure Event Standard for the added HFEs have not                                                    (HFE) credited in been performed.                                                                          the FPRA.
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Elements      Application A focused scope The added HFEs need to be processed                                                peer review was HRA-B3    using the same methods that were                                                      performed in June (continued) employed to develop the HFE for the                                                  2014 evaluating internals event PRA.                                                                  the revised HRA process.
Fire PRA Peer Review Finding HRA-C1-01:
A review of the fire HFE Evaluation and the recovery rules revealed a discrepancy for one HFE. As shown in the ANO-1-Fire HFE Evaluation spreadsheet, the HFE EHF1DGCRKP had a value of 4.98E-03 in the internal events PRA for ANO-1 with a new value of 2.99E-02 calculated based on This issue the fire conditions. The calculation was was resolved reviewed and found to match the fire HEP    The value Z1EHFDGCRK was and therefore process. However, when reviewing the        corrected in the FPRA rule recovery file has no impact recovery rule file, Alrul4p00.txt, the      (Alrul4p00_FIRE.txt). The recovery rule  Original in STI change N/A  Resolved  HRA-C1    replacement event for EHF1DGCRKP,            file was thoroughly reviewed during the  disposition evaluations Z1EHFDGCRK, was found to have the            HEP document update to ensure the        remains valid.
performed in original value 4.98E-03. It was determined  correct values are applied to the events accordance that the error was a result of an error when during recovery.
with the copying the values from one file to another.
SFCP.
One error was found in a small sampling so the extent of condition may be larger so may impact the results.
Correct the value for Z1EHFDGCRK in Alrul4p00.txt and then review the other "single replacement" values against the new values in the ANO-1-Fire HFE Evaluation spreadsheet.
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Elements      Application Fire PRA Peer Review Finding HRA-D2-01:
Multiple recovery actions were inserted into the model via the AlteredEvents file.        The HRA events have been removed Screening values were used for all of these  from the Altered Events table, except events so none of them accounted for        for actions that are set to TRUE in a fire This issue relevant fire-related effects, including any scenario where fire prohibits the was resolved effects that may preclude a recovery action  operator action. The revised FPRA and therefore or alter the manner in which it is          HRA analysis (PRA-A1- 05-008 -
See new          has no impact accomplished. The values used may be        ANO-1 Fire PRA Human Failure disposition      in STI change N/A  Resolved  HRA-D2    conservative or non-conservative so it is    Events) provides sufficient detail to response for HRA- evaluations not possible to fully assess the true impact meet all HRA-D2 supporting B3-01            performed in of these recovery actions.                  requirements. These changes include accordance ANO-1 plans to determine which of these      new combinations of operator actions with the recovery actions need to be retained after  for dependency. All events used in the SFCP.
the NFPA 805 Change Evaluation. Once        FPRA have been developed and the HFEs to be retained are identified, they documented using the same methods need to be fully defined and quantified in  used for the internal events HEPs.
accordance with the process used for all the other HFEs.
Fire PRA Peer Review Finding HRA-E1-01:
In the FRANC Altered events file, there are a number of basic events with replacement values of 0.1. These replacement values represent screening values for Operator                                                                  This issue Recovery Actions to recover the faulted      The HRA events have been removed was resolved basis event. This is the only place these    from the Altered Events table, except and therefore operator actions show up, they are not    for actions that are set to TRUE in a fire See new          has no impact proceduralized at this time and they are not scenario where fire prohibits the disposition      in STI change N/A  Resolved  HRA-E1    documented or evaluated beyond the          operator action. As discusses in the response for HRA- evaluations screening evaluation. At this point in time, response to HRA-B3-01 and HRA-D2-B3-01            performed in these new actions are not proceduralized    01, these actions are evaluated in detail accordance and are considered to be recovery actions. using the same HRA methodology used with the These actions still need to be evaluated for in internal events model.
SFCP.
significance. Entergy has indicated that once these actions have been evaluated, the important ones will be incorporated into procedures and quantified in accordance with their standard process.
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Elements    Application The other actions will be removed. Inclusion of undocumented, unevaluated operator actions in the models can impact the results.
Before the fire PRA can be used for HRA-E1    applications beyond NFPA 805, the (continued) "implied human actions" need to be documented and incorporated in operating procedures. Each such action needs to be clearly identified. Any actions that are not proceduralized need to be removed from the model.
Fire PRA Peer Review Finding HRA-E1-02:
Report 0247060006.03-U1, Rev. 0 documents the HRA for the ANO-1 Fire PRA. Section 3 documents the assumptions used in the Fire PRA HRA.
This section contains a total of 2 assumptions. However, a text search of the report on assume found five additional  Section 5 of the ANO-1 Fire PRA                        This issue assumptions buried in the text. Another      Human Failure Events report (PRA-A1-                  was resolved text search on could and a text search on  05-008) has been expanded and now                      and therefore may yielded another three instances of      includes all relevant HRA modeling                    has no impact what appeared to be assumptions. This is                                              Original assumptions. In addition to the general                in STI change N/A  Resolved  HRA-E1    considered to be a good indication that not                                          disposition assumptions included in Section 5 of                  evaluations all assumptions have been documented.                                                remains valid.
calculation PRA-A1-05-008, each of the                performed in While capturing all assumptions into a        detailed post-fire HRA events has                      accordance common location may not have a                assumptions included in the associated                with the significant impact on the base model, there  evaluation.                                            SFCP.
is a concern for future applications. One step in performing a risk-informed application is to review the assumptions to determine if any of them could impact the application and, if so, what would need to be done to compensate for the assumption if it is nonconservative with respect to the application.
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Elements    Application Review Report 0247060006.03-U1, Rev. 0 to identify additional assumptions and capture them in Section 3. The definition of "assumption' used for the search should be HRA-E1 relatively broad so as to capture as may (continued) potential assumptions as possible. It is easier to disposition a trivial assumption than it is to address a significant assumption that was not identified as such.
Fire PRA Peer Review Finding FQ-A4-01:
The uncertainty interval on CDF results This issue was not estimated as required by QU-E3      Calculation PRA-A1-05-006 - ANO-1 was resolved (LE-F3). An uncertainty analysis has been    Fire PRA Uncertainty/Sensitivity and therefore performed to identify and qualify specific  Analysis provides a Monte Carlo has no impact areas of uncertainty. This meets the        evaluation of uncertainty for both the    Original in STI change N/A  Resolved    FQ-A4    internal event requirement for QU-E1,        FPRA Core Damage Frequency and the        disposition evaluations QU-E2, and QU-E4.                            Large Early Release Frequency. The        remains valid.
performed in Determine an uncertainty interval based      listed uncertainty analysis satisfies the accordance upon the model uncertainties identified in  listed Standard requirements (QU-E1, with the QU-E1 and E2. Provide basis for any non-    QU-E2, and QU-E3).
SFCP.
applicability of any of the requirements under these sections in Part 2.
Fire PRA Peer Review Finding FQ-A4-02:      The scenario ignition frequencies are This issue A spot check of scenario ignition            calculated in Attachment D of the Fire was resolved frequencies documented in the FRANC          Scenarios report (PRA-A1-05-004).
and therefore model revealed several errors in the        The calculated scenario ignition has no impact calculations. Based on the number of        frequencies have been verified to be      Original in STI change N/A  Resolved    FQ-A4    errors found and the lack of documentation  consistent with the calculations          disposition evaluations for scenario frequency calculations, this    performed for the zone frequencies in    remains valid.
performed in indicates a potentially systemic problem    the Fire Probabilistic Risk Assessment accordance with the scenario ignition frequencies.      Plant Partitioning and Fire Ignition with the Review and recalculate scenario ignition    Frequency Development report (CALC-SFCP.
frequencies.                                08-E-0016-01).
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Elements    Application Fire PRA Peer Review Finding FQ-A4-03:
Section 7.4 of Appendix D (MSO Expert Panel Review and Disposition of Open Items) in Report 0247060006.02-U1                                                                  This issue describes changes made to the PRA model    The modeling error has been corrected.                  was resolved to account for the potential of an MSO    The new event (RHF1RCPSXP) has a                        and therefore causing a spurious spray event. A review  default value of 1.0. It is only changed                has no impact Original of the model shows that the AND gate      if changed in the altered events table                  in STI change N/A  Resolved  FQ-A4                                                                                        disposition FIRE027 that represents this scenario      when it is relevant to a specific case.                  evaluations remains valid.
includes an event (RHF1RCPSXP) that is    With the default value of 1.0, it is does                performed in set to 0.0. This will prevent the MSO      not disrupt quantification of the other                  accordance scenario from being quantified. The model  logic in the AND gate for other cases.                  with the does not accurately quantify an MSO                                                                SFCP.
scenario due to a modeling error.
Correct the model and review for other potential similar errors.
Fire PRA Peer Review Finding FQ-D1-01:
The Fire PRA LERF model is based upon the internal events LERF model. The LERF  F&Os relating to ANO-1 ISLOCA model uses the Fire PRA plant response    treatment have all been resolved.
model. The frequency for fire-induced      A revision to the ISLOCA fault tree was LERF is quantified. However, F&Os for      performed following the peer review.                    This issue element "LE" and other elements were      This revision included an update to the                  was resolved identified in the ANO-1 RG 1.200 peer      internal events model (and                              and therefore review of the internal events model,      subsequently the FPRA model).                            has no impact Original performed in August 2009. These F&Os                                                                in STI change N/A  Resolved  FQ-D1                                              The evaluation of containment isolation  disposition have not been addressed and limit the                                                              evaluations paths (potential LERF contribution via    remains valid.
LERF modeling capability of the Fire PRA                                                            performed in breaches in containment) are model. Also, comprehensive screening of                                                            accordance documented in Appendix G of the Interfacing Systems LOCA (ISLOCA) and                                                              with the Component and Cable Selection Report other significant containment isolation                                                            SFCP.
(PRA-A1-05-003).
paths for fire scenarios has not been performed. The frequency of different      See Table 2 for ISLOCA/LERF findings containment failure modes leading to large additional details.
early release is needed for fire-induced LERF.
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Elements    Application Resolve internal events model F&Os for element "LE." Comprehensive screening FQ-D1 of Interfacing Systems LOCA (ISLOCA)
(continued) and other potential significant containment failure paths is also needed.
Fire PRA Peer Review Finding FQ-E1-01:      Since the ANO-1 FPRA Peer Review, The ANO-1 Fire PRA results are very        ANO has made several refinements conservative for CDF and LERF. There        within the reviewed methodology to are several important scenarios that are    remove some conservatism and reduce driving the results that have not had      overall CDF and LERF. These                              This issue detailed modeling performed to reduce the  refinements include:                                    was resolved conservatisms. These conservative results        developing more detailed fire                    and therefore may mask other important contributors to          scenarios                                        has no impact the fire risk. This SR requires that                                                See new response in STI change N/A  Resolved  FQ-E1                                                      refining the components failed significant contributors be identified in                                          to FSS-C1-01. evaluations within the scenario accordance with HLR-QU-D. HLR-QU-D6                                                                  performed in requires that significant contributors be        refining the fire HRA events and                  accordance identified and HLR-QU-D7 requires review          JHEPs.                                            with the of important components and basic events    Detailed fire modeling has not been                      SFCP.
to determine that they make logical sense. applied to ANO fire scenarios based on This is not possible with overly            the limited benefit in CDF and LERF conservative scenario models.              reduction given conservative input Update model to remove conservatisms.      parameters.
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Elements    Application The ANO-1 Fire PRA Summary Report Fire PRA Peer Review Finding FQ-E1-02:    (PRA-A1-05-005) has been updated to include additional result details.
The ANO-1 Fire PRA Summary Report          Attachments D, E, F, G, & H have been (ERIN Report 0247060006.06-U1, Rev. 1,    added to provide additional details on 9/30/09) Appendices A and B provide the  results.
quantification results for CDF and LERF.
Appendix C presents INSIGHTS /                  Appendix D - Uncertainty and                      This issue RECOMMENDATIONS (DOMINANT RISK                    Sensitivity Matrix                                was resolved CONTRIBUTORS). The significant                  Appendix E - Cutsets                              and therefore contributors to LERF were not well                Comprising the Top 90% of CDF                    has no impact Original identified and results were not clearly                                                            in STI change N/A  Resolved  FQ-E1                                                    Appendix F - CDF Importances      disposition traced to inputs and assumptions in the                                                            evaluations remains valid.
Fire PRA. Therefore the SR is not met.            Report                                            performed in The fire-induced LERF quantification                                                                accordance Appendix G - Cutsets results are not reviewed sufficiently to                                                            with the Comprising the Top 90% of identify significant contributors to LERF.                                                          SFCP.
LERF Presentation of dominant LERF risk contributors in Appendix C should be            Appendix H - LERF Importances expanded to fully discuss all dominant            Report contributors and their basis in the inputs Section 4.0 of the Summary Report has and assumptions made in the Fire PRA.      been updated to explain results, insights, and dominant risk contributors.
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Elements      Application Section 4.2 of the ANO-1 Fire PRA Fire PRA Peer Review Finding FQ-F1-01:      Summary Report (PRA-A1-05-005) discusses the limitations of the PRA There is no discussion of quantification software used (CAFTA, FORTE,              The original process limitations as required in QU-F5, or                                                              This issue FRANC). Quantitative results and          disposition significance definitions (basic event,                                                                    was resolved insights for risk significant sequences    remains valid with cutsets, and accident sequences) as                                                                        and therefore are also provided in the scenario report. the exception that required by QU-F6. A discussion of the                                                                    has no impact Significance is defined in Section 4.1 of  quantitative quantification process limitations is                                                                      in STI change N/A  Resolved  FQ-F1                                                the report as all scenarios included in    results and required by QU-F5 by reference through                                                                    evaluations 90% of the total Fire CDF/LERF.            insights for risk FQ-F1. Also, quantitative definitions for                                                                  performed in Tables in the report list risk significant significant significant basic events, cut sets, and                                                                    accordance scenarios (4-1 for CDF and 4-2 for        sequences are not accident sequences are required by                                                                        with the LERF), cutsets (Appendix E for CDF        provided in the QU-F6.                                                                                                    SFCP.
and Appendix G for LERF), and basic        scenario report.
Provide the required discussions and        event importance measures definitions and document.                    (Appendix F for CDF and Appendix H for LERF).
Fire PRA Peer Review Finding UNC-A1-01:
The uncertainty interval on CDF results was not estimated as required by QU-E3 Uncertainty intervals (ones that meet the                    This issue (LE-F3). An uncertainty analysis has been criteria of QU-E1, E2, and E3) have been                      was resolved performed to identify and qualify specific developed for both CDF and LERF and                          and therefore areas of uncertainty. This meets the are documented in PRA-A1-05-006 -                            has no impact internal event requirement for HLR-QU-E1,                                              Original ANO-1 Fire PRA Uncertainty/Sensitivity disposition            in STI change N/A  Resolved  UNC-A1    E2, and E4. The uncertainty interval on Analysis.                                                    evaluations CDF results was not estimated as required                                              remains valid.
performed in by QU-E3 (LE-F3).                            The results of the uncertainty evaluation                    accordance Determine an uncertainty interval based      are also presented in Appendix D of the                      with the upon the model uncertainties identified in  Summary Report (PRA-A1-05-005).                              SFCP.
QU-E1 and E2. Provide basis for any non-applicability of any of the requirements under these sections in Part 2.
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Elements    Application Fire PRA Peer Review Finding FSS-G3-01:
This issue There is no documented basis for the use The updated Multi-Compartment and                    was resolved of screening criteria value of 5E-7/yr. In Hot Gas Layer analysis no longer uses                and therefore addition, given that the threshold value is screening criteria for HGL (ZOI) and                  has no impact 1E-7, it is not documented in the report if                                            Original MCA. This is documented in PRA-A1-                    in STI change N/A  Resolved  FSS-G3    the screening process has considered                                                    disposition 05-009 -ANO-1 Multi-Compartment/ Hot                  evaluations cumulative risk.                                                                        remains valid.
Gas Layer Analysis. All zones now                    performed in Provide a justification for the screening        include a HGL and multiple MCA                        accordance value of 5E-7. The report should discuss        scenarios.                                            with the how the screening process deals with                                                                  SFCP.
cumulative risk.
Fire PRA Peer Review Finding FSS-G3-04:
The assumption of only two cable trays as intervening combustibles may not be This issue conservative or realistic (i.e. reflect as built was resolved conditions) for all the PAU's. Given that and therefore the multi compartment analysis relies Walk-downs to identify scenarios with                has no impact heavily on screening due to hot gas layer                                              Original greater than 2 trays were performed                  in STI change N/A  Resolved  FSS-G3    conditions in the PAU, accurate intervening                                            disposition and the results have been incorporated                evaluations combustible input parameters to the fire                                                remains valid.
into the FPRA (PRA-A1-05-009).                        performed in modeling analysis is necessary.
accordance Conduct walkdowns for identifying the                                                                  with the correct package of intervening                                                                        SFCP.
combustibles to use as input to the fire modeling analysis. Walkdowns should be documented.
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Elements          Application The original disposition to this F&O is not valid for the current modeling. The following response better depicts the current Fire PRA Fire PRA Peer Review Finding FSS-G3-05:                                                model.
The use of the manual suppression                                                      The manual non-constant for electrical fires appears to be                                            suppression used for the calculation of all the manual                                              probability (NSPms) suppression failure probabilities. The                                                  is calculated using a manual suppression constant should be                                                  convolution process which applies the applied depending on the specific ignition                                                                    This issue above non-source/fire that characterizes the fire      The updated ANO-1 Multi-Compartment      suppression curve was resolved scenario. In addition, it is not clear in the Analysis (PRA-A1-05-009) contains a      to each bin of the    and therefore calculation which time input is used for      discussion of how manual suppression      applicable            has no impact determining the manual suppression            probabilities are determined. This        NUREG/CR-6850,        in STI change N/A  Resolved  FSS-G3 probability values (i.e. at what time hot gas discussion contains the technical details Appendix E heat      evaluations layer is reached?). Based on the response    (including justification). The .012      release rates (HRR),  performed in to a question submitted during the peer      non-suppression curve, when applied,      with the time to hot  accordance review process, there are some curves        is applied in a bounding condition.      gas layer varying    with the (e.g. the high energy arcing fault) that also                                          based on the heat    SFCP.
could apply to specific PAU's that were not                                            release rate considered.                                                                            associated with each bin. The Ensure that the suppression curve selected mean suppression is bounding and document a technical                                                    rate from Table 14-1 justification for the selection to be used in                                          of NUREG/CR-6850 the screening process.                                                                  Supplement 1 for the appropriate fire type is used as opposed to the
                                                                                                                    .0102 bounding value discussed in the original disposition to this F&O.
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Elements    Application Fire PRA Peer Review Finding FSS-G3-06:
There are numerous statements in Table 3-1 under "final disposition" that lack technical justification. Examples include (this F/O is not limited to these examples only):
                                "The 0.001 applied is a very conservative factor applied to zones over 353 cu ft. A more appropriate NSPms for this scenario is 1E-04                                                              This issue which results in screening the          Table 3-1 (the table in question) has                  was resolved scenario for MCA impacts" -            been eliminated. It was removed via                    and therefore Question: Where does 0.001 come        the change in screening approach (to                  has no impact Original from? Why it is considered              address F&O FSS-G3-01). All zones                      in STI change N/A  Resolved  FSS-G3                                                                                          disposition conservative? Where does 1E-4          now have an HGL scenario and multiple                  evaluations remains valid.
come from?                              MCA scenarios and the associated                      performed in Why is a factor from FAQ 044 for        quantification is incorporated into the                accordance main feed water pumps (MFWPs) to        FPRA model.                                            with the oil tank rooms used? Are there any                                                            SFCP.
other ignition sources other than the pumps in these areas?
Given the large volume of the turbine building, no hot gas layer would be able to form which would preclude the MCA impacts - Question: Is this true for MFWP and large turbine generator (TG) fires? Can we preclude HGL scenarios for these ignition sources?
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Elements  Application The only adjacent zone connected through a door is 2200-MM which will not be impacted due its large volume.
Other adjacent zones can use the next worse barrier failure probability for dampers, 0.0027, which lowers the Pmca to 9.94E Comment: The resulting screening value is barely below the screening criteria. If we are selecting probabilities from a table without considering what is in the boundary in terms of seals and dampers, proper justification for the probabilities are needed.
Crediting a 0.02 NSP for emergency diesel generator (EDG) oil fires from FSS-G3          Appendix E screens the scenario (continued)      without crediting manual suppression
                                  - Question: Where is this 0.02 coming from? Where is the reference for the fixed system credited?
There are tray combustibles within the zone; however they are located 10 ft or more above the floor elevation and would not be impacted by a fire in this zone. Screen the scenario for MCA impacts - Question: Why does the statement "would not be impacted by a fire in this zone" apply to these specifically?
Clarify these statements in a way that simplify reviews and future updates of this calculation. Some of these will require clear technical justifications.
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Elements    Application Fire PRA Peer Review Finding FSS-G4-01:
It is not documented in the multi compartment report how the requirements of this SR were addressed. For example, Walk-downs of non-NRC or Insurance the standard requires to "CONFIRM that commitment fire barriers have been the allowed credit is consistent with the fire-performed to document the basis for resistance rating as demonstrated by                                                                    This issue credit taken for fire zone boundaries.
conformance to applicable test standards".                                                              was resolved Quantification of MCA probability has There is no evidence that this confirmation                                                              and therefore conservatively used the door failure has been conducted. Without a systematic                                                                has no impact probability from NUREG/CR-6850, process for identifying barrier types                                                    See response to in STI change N/A  Resolved  FSS-G4                                                    Table 11-3, as the boundary failure between physical analysis units, ANO-1 will                                              PP-C3-01.      evaluations mechanism for all zones without need to ensure that addressing this SR will                                                              performed in openings to adjacent fire zones. For account for all the different barrier types                                                              accordance zones with openings to adjacent zones, (walls, barriers, spatial separations, doors,                                                            with the the boundary failure probability was set etc.). See F/O PP-B3-01.                                                                                SFCP.
to 1.0 and the volume of the combined A possible resolution is to list the types of  zones was used for assessing the time barriers between adjacent PAU's for            to HGL formation (PRA-A1-05-009).
determining which probabilities are applicable and document if the generic values in Table 11-3 of NUREG/CR-6850 are bounding for ANO-1.
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Elements    Application Fire PRA Peer Review Finding FSS-G5-01:
It is not documented in the multi compartment report how the requirements of this SR were addressed. For example,        Walk-downs of non-NRC or Insurance the standard requires to "QUANTIFY the          commitment fire barriers have been effectiveness, reliability, and availability of performed to document the basis for the active fire barrier element". There is no                                                            This issue credit taken for fire zone boundaries.
evidence that this confirmation has been                                                                was resolved Quantification of MCA probability has conducted. Without a systematic process                                                                  and therefore conservatively used the door failure for identifying barrier types between                                                                    has no impact probability from NUREG/CR-6850, physical analysis units, ANO-1 will need to                                              See response to in STI change N/A  Resolved  FSS-G5                                                    Table 11-3, as the boundary failure ensure that addressing this SR will account                                              PP-C3-01.      evaluations mechanism for all zones without for all the different barrier types (walls,                                                              performed in openings to adjacent fire zones. For barriers, spatial separations, doors, etc.).                                                            accordance zones with openings to adjacent zones, See F/O PP-B3-01.                                                                                        with the the boundary failure probability was set SFCP.
A possible resolution is to list the types of  to 1.0 and the volume of the combined barriers between adjacent PAU's for            zones was used for assessing the time determining which probabilities are            to HGL formation (PRA-A1-05-009).
applicable and document if the generic values in Table 11-3 of NUREG/CR-6850 are bounding for ANO-1.
Fire PRA Peer Review Finding FS HR-G3-01:
Observation: Detailed analysis of HPI-HFC-FO-CRSPR-EF and HPI-HFC-FO-                                                                          This issue HR-G3 CRSPR-IF are two examples where the            HFEs HPI-HFC-FO-CRSPR-EF/-IF                            was resolved (Finding time available is less than the time            were re-examined consistent with the                    and therefore not required.                                      detailed quantification method                          has no impact submitted discussed in Section 7 of PRA-A1                    in STI change A1-5136 Resolved  in original Basis for Significance: The HCR/ORE                                                      N/A 015 R0 and timing information was                        evaluations Att. V of  quantification for these HFEs are non-          updated such that the time available is                  performed in the ANO-1    conservative and the HEP should be 1.0.        greater than the time required for the                  accordance NFPA 805 Possible Resolution: Set HEPs for HPI-          operator action.                                        with the LAR)
HFC-FO-CRSPR-EF and HPI-HFC-FO-                                                                          SFCP.
CRSPR-IF to 1.0 or review and update timing information such that time available is greater than time required.
to 1CAN031801 Page 92 of 102 Changes to  Importance MCR            Applicable Status                            Finding/Observation                              Disposition                    Modeling        to Number              SR(s)
Elements  Application Fire PRA Peer Review Finding FS HR-G3-02:
Observation: All assumptions that impact feasibility of operator actions need to be validated before the HEPs can be applied, specifically:
Assumptions regarding availability of instrumentation credited for diagnosis        The new fire specific HFEs developed should be verified available in the fire      in PRA-A1-05-015 R0 were reviewed to scenarios where the HFE is credited. This    identify assumptions related to includes instrumentation credited as          feasibility that would need to be diverse to instrumentation that is rendered  addressed for NFPA 805; none were unavailable due to fire. Note that these      identified. [NOTE: feasibility analysis assumptions may be implicit in analyses      was performed for each HFE that had where cues and indications are stated        been identified as risk-significant on the without any statements regarding              basis of the Fussell-Vesely importance This issue HR-G3    availability in the fire scenarios where the  measure. The fire PRA does not take was resolved (Finding  HFE is credited.                              credit for non-risk significant HFEs; and therefore not                                                  these full power internal events (FPIE)
Related to the above are assumptions                                                                  has no impact submitted                                                HEPs were set to 1.0.]
made with regards to explicit modeling of                                                              in STI change A1-5137 Resolved  in original                                                                                          N/A instrumentation. Assumptions regarding        PRA-A1-05-003 R2, Appendix F                            evaluations Att. V of procedure changes need to be validated.      (Pages F-34 through F-51) show screen                    performed in the ANO-1 Basis for Significance: The reliability of    shots of the HRA cue instrumentation                    accordance NFPA 805 operator actions can only be assessed for    added. The model change events log                      with the LAR) feasible operator actions. If an operator    (Appendix F) provides the listing of the                SFCP.
action is not feasible, it cannot be reliable changes to the fault tree for the and the HEP should be 1.0 (refer to          FIRE_HRA_1 through FIRE_HRA_18.
NUREG/CR-1921 Sections 3.5 and 4.3 for        The original HRA was contained in additional discussion). Risk metrics can be  PRA-A1-05-007 R0 and PRA-A1                              underestimated if feasibility is assumed      008 R1 and combined into a single HRA where operator actions cannot be shown to    report PRA-A1-05-015 R1 after the be feasible.                                  Peer review and NFPA-805 RAI Possible Resolution: Ensure that all          responses.
assumptions regarding feasibility (whether implicit or explicit) are systematically identified and addressed. It will be necessary to demonstrate this to support the intended NFPA-805 license change request.
to 1CAN031801 Page 93 of 102 Changes to  Importance MCR            Applicable Status                            Finding/Observation                                Disposition                  Modeling        to Number              SR(s)
Elements  Application Fire PRA Peer Review Finding FS HR-G7-01:
Based on a review of Figure 5 in PRA-A2-01-003S03, the dependency approach does not consider availability of resources, which can be important for fire PRA.
Basis for Significance:
The number of operators required to perform actions implied by a combination of                                                            This issue HR-G7 HFEs in a cutset may exceed the number          The dependency analysis performed for                was resolved (Finding of operators available to perform these          the current analysis (PRA-A1-05-015                  and therefore not actions. As such some actions will not be        R0) utilizes the EPRI HRA Calculator,                has no impact submitted able to be performed and the HEPs should        rather than the HRA Toolbox, and                      in STI change A1-5139 Resolved  in original                                                                                          N/A be 1.0. For standard internal events PRAs,      therefore addresses the peer review                  evaluations Att. V of this is not typically an issue, but for fire and concerns regarding the assessment of                  performed in the ANO-1 other spatial events, the need for more          dependency factors such as availability              accordance NFPA 805 operator actions may produce cutsets with        of resources.                                        with the LAR)
HFEs that are inappropriately credited.                                                                SFCP.
Possible Resolution: Review the combinations and determine resources required versus resources available.
Justify credit given tor HFEs where resources required < resources available e.g. for long enough time windows, additional manpower will be on site to assist.
to 1CAN031801 Page 94 of 102 Table 5 List of SRs Assessed as CC-I in the ANO-1 Fire PRA Model Importance Changes to Modeling Elements to SR            Topic                                          Status                                                                            to Reflect New Guidelines Application The original disposition for this F&O does not depict the modeling used in the current MOR.
The following information was obtained from the ANO-1 SE (ML16223A481). In PRA RAI 01.c (Reference 21), the NRC staff requested that the licensee explain its approach and whether the split fraction referred to in the Capability Category 1 is acceptable for the application. While the      resolution to FSS-C1-01 implied results are conservative, they are not significantly more so than the    that the "Panel Factors method" also conservative results of more detailed fire modeling.                (Reference 105), not accepted by      This issue Additionally, some multi-point heat release rate analysis is applied. NRC was used. In its response to      has no impact Section 7.1 of the Fire Scenarios Report outlines the use of a multi-    PRA RAI 01.c (Reference 10), the      in STI change Use of Multi-Point                                                                          licensee clarified that that the point treatment for vented panels based on a split fraction developed                                          evaluations FSS-C1    Heat Release Rate                                                                          "Panel Factors" method was not from the EPRI Fire Events Database. This split fraction specifies the                                          performed in Treatment                                                                                  used but rather generic fire fraction of fires impacting only the ignition source panel versus those                                        accordance fires which impact targets within the zone of influence of the panels. modeling treatments (GFMTs)          with the were used to define a three-point    SFCP.
Section 16 of the Fire Scenarios Report discusses the use of generic    fire model to meet SR FSS-C1.).
fire modeling versus detailed fire modeling and justifies the ANO        The licensee explained that the approach for the FPRA application.                                      first fire in the three-point fire treatment is a non-severe fire in which the source panel and the cables terminating at the source panel are damaged but the nearest target is not damaged.
The second fire in the three-point fire treatment is a severe fire in which all targets within the 98th percentile zone of influence (ZOI) are impacted.
to 1CAN031801 Page 95 of 102 Importance Changes to Modeling Elements to SR            Topic                                            Status                                                                              to Reflect New Guidelines Application The third fire of the three-point fire treatment is a fire that results in a hot gas layer (HGL) exceeding a 80 degree Centigrade (&deg;C) criterion in which all targets in the fire zone are conservatively assumed to be damaged. These three fire models are discussed in the licensee's response to FM RAI 01.f discussed in SE FSS-C1 Section 3.4.2.3. The NRC staff (continued) concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that it used a multiple-point fire intensity and duration model encompassing low likelihood but risk-contributing fire events consistent with the requirements in the PRA standard.
In limited PAUs, ANO credits non-rated barriers (spatial separation) and does not full meet Capability Category II. The impacts of these barriers are evaluated (no credit for non-rated barrier) in the scenario development and MCA.
SR PP-B2 references NUREG/CR-6850, Chapter 1, for the acceptable                                                This issue criteria for justifying non-rated fire barriers. NUREG/CR-6850                                                  has no impact discusses the use of fire compartments as a well-defined enclosed                                              in STI change PP-B2    Plant Partitioning -
room, not necessarily with fire barriers. ANO references the FHA as                                            evaluations (Finding  Use of Non-Rated                                                                              Original disposition remains valid.
a starting point for plant partitioning and all barriers (both rated and                                        performed in PP-B2-01)  Fire Barriers non-rated are defined in the FHA). The Plant Partitioning Task                                                  accordance (CALC-08-E-0016-01, R0) assumes that fire protection features will be                                            with the effective at containing a fire under most conditions. Fire protection                                            SFCP.
features include fire-rated barriers, non-fire-rated barriers, active features such as water curtains, and in some cases spatial separation.
The potential failure of a credited partitioning feature is addressed in the MCA.
to 1CAN031801 Page 96 of 102 Importance Changes to Modeling Elements to SR            Topic                                            Status                                                                          to Reflect New Guidelines Application The ANO FHA does not include any partitioning features, such as PP-B2 partial height walls, that are discussed in NUREG/CR-6850 as barriers (Finding that should not be credited. The adequacy of the fire barriers is PP-B2-01) explicitly reviewed as part of the Multi-Compartment and Hot Gas (continued)
Layer Analysis calculation (PRA-A1-05-009).
The original F&O disposition is not applicable. The following Spatial separation is used in two areas at the site (four total PAUs):  response better depicts the        This issue the fuel handling area and the turbine deck. Both are very large areas  current Fire PRA model.            has no impact and the development of a hot gas layer is not credited. The spatial    The current analysis considers      in STI change Plant Partitioning -
separation distance is also sufficiently large such that no fixed or    spatial separation in a very few    evaluations PP-B3-01  Use of Spatial transient sources are capable of impacting the area beyond the          areas. In each of these areas, the  performed in Separation separation. While this only meets Capability Category I for            MCA/HGL analysis assumes no        accordance partitioning, the follow up tasks (fire scenarios and MCA) show it has  barrier failure probability.        with the no impact on FPRA results or conclusions.                              Therefore, the significance of the  SFCP.
spatial separation is assessed in the analysis results.
to 1CAN031801 Page 97 of 102 Importance Changes to Modeling Elements to SR              Topic                                          Status                                                                              to Reflect New Guidelines Application The determination of MCR abandonment scenario was updated to incorporate the following elements into the analysis; 1) determination of the frequency of abandonment using the multiple ignition sources within SR was given a CC-I due to the assumptions associated with the MCR        the control room, 2) fire protection abandonment Scenario.                                                    features, room ventilation and room geometry are considered.
The Main Control Room Abandonment analysis uses a bounding                3) NSP values were selected from approach to characterize fire risk contribution. Section 3.2 of the Fire  Calc ANO1-FP-09-00011 Potential fire        Scenarios Report (0247-06-0006.05-U1) details MCR abandonment                                                  This issue assuming multi-cable bundle fire scenarios leading to  treatment. The scenario is 129-F, Scenario A. A CCDP of 0.1 is used                                            has no impact spreads, 4) CCDP was calculated the MCR              based on an adequate evaluation of Appendix R, III.L requirements.                                            in STI change based upon the failures modeled abandonment.          The evaluation is based on adequate alternate shutdown procedures,                                            evaluations FSS-B2                                                                                                    to occur from a fire inside the (Issue not submitted  validation of timing and manual action feasibility. Calculation ANO1-    control room, 5) credit for          performed in in original Att. V of FP 00011, Rev. 2, evaluates the abandonment times for the Unit 1      mitigating operator actions outside  accordance the ANO-1            MCR. These arguments bound the fire risk contribution for MCR            the control room. The                with the NFPA 805 LAR)        abandonment.                                                              consideration of these elements in  SFCP.
Additionally, an incorrect calculation reference (Reference 3 in          determining the CCDP of the Section 8) of the Fire Scenarios Report (0247-06-0006.05-U1, Rev. 0). MCR abandonment scenario The calculation reference is ANO2-FP-09-00011. The correct                addressed the bounding reference should be ANO1-FP-09-00011.                                    assumptions identified during the peer review and provides a realistic assessment of the MCR abandonment scenario.
Additionally, the incorrect reference was corrected in subsequent revisions to the Scenario documentation.
to 1CAN031801 Page 98 of 102 Importance Changes to Modeling Elements to SR              Topic                                            Status                                                                              to Reflect New Guidelines Application The Uncertainty analysis for the ANO-1 Fire PRA model includes propagation of uncertainty for circuit failure model probabilities.
The Hot Short Probabilities (HSPs) were incorporated using a similar approach to the fire ignition Characterize the                                                                                frequencies. A type code was uncertainty          The summary report, Appendix D for task 10 characterizes the              developed for each HSP utilized in associated with the  uncertainty associated with method employed in determining failure        the ANO-1 model. The mean and        This issue applied conditional  probabilities. The conclusion of this report is that the "application of  its associated variance were          has no impact failure probability  circuit failure probabilities is considered to have minimal impact on the  assigned based on the information    in STI change assigned for fire-    results." Though it may be that a detailed analysis technique was          in NUREG/CR-7150. This was            evaluations CF-A2    induced circuit      followed for dominant scenarios, these failures are still in the dominant  done to correlate uncertainty in      performed in failures              sequences. The accuracy and uncertainty associated with these              the HSPs associated with each        accordance (Issue not submitted  values would have a significant impact on the results.                    fire scenario. This approach          with the in original Att. V of Characterize the uncertainty with respect to how the method employed      avoided the inappropriate            SFCP.
the ANO-1            could introduce uncertainty into the final results.                        assumption that the scenario NFPA 805 LAR)                                                                                    frequencies are independent of each other. Latin Hypercube sampling was performed to propagate parametric uncertainties through the ANO-1 FPRA model to generate probability distributions for ANO-1 fire CDF and LERF.
to 1CAN031801 Page 99 of 102 Importance Changes to Modeling Elements to SR              Topic                                          Status                                                                                  to Reflect New Guidelines Application The original internal events HEPs that were carried over to the FPRA were evaluated in detail through operator interviews and simulator exercise experience as documented in Appendix B of the HRA analysis. Fire specific information used to update these HFEs for the Fire PRA was Talk through (i.e.,                                                                              verified with PRA staff with review in detail) with                                                                          operations and training plant operations and                                                                            experience.
training personnel the procedures and    ANO-1 reviewed the required cues for each identified action against      As part of detailed analysis of sequence events to    the Appendix R protected instrumentation to confirm that adequate        specific events for fire, a feasibility confirm that          cues would be available. The results of these reviews are                assessment consistent with the This issue interpretation of the  documented in Attachment A of the report. Attachment E contains the      guidance in NUREG-1921 was has no impact procedures relevant    results of the simulator review study to confirm the availability of the  performed. FPIE HFEs were in STI change to actions identified  Appendix R cues and the likely actions. These discussions were at a      identified as risk-significant on the evaluations HRA-A4    in SRs HRA-A1,        very high/general level and did not address specific procedures.          basis of the F-V importance performed in HRA-A2, and HRA-                                                                                measure. These risk-significant ANO should strengthen the discussions, especially as they pertain to                                              accordance A3 is consistent with                                                                            FPIE HFEs formed the basis for the conclusion that the operators will not take action based on a single                                          with the plant operational                                                                                the development of new fire-indication.                                                                                                      SFCP.
and training                                                                                    specific HFEs, which were ANO should also explicitly identify which specific Appendix R            evaluated for feasibility.
practices.
instruments are to be credited for each identified human action.
(Issue not submitted                                                                            The following areas were evaluated in original Att. V of                                                                            to ensure all actions were feasible, the ANO-1                                                                                        and if any action was determined NFPA 805 LAR)                                                                                    to be non-feasible, a value of 1.0 was assigned to the operator action. The following parameters were evaluated for feasibility:
procedures and training, available indications and cues, equipment functionality and accessibility, adequate time available to perform the action, environmental factors, communications, and staffing.
to 1CAN031801 Page 100 of 102 Importance Changes to Modeling Elements to SR              Topic                                          Status                                                                            to Reflect New Guidelines Application Cues and indications are necessary because all required operator actions are preceded by them. Without cues or indications, the operators have no prompts that an action is required and, hence, no operator action can be credited. The analysis must evaluate whether or not the instruments and indications HRA-A4 needed for diagnosis are affected (continued) by the fire. The current HRA analysis provides information that indicates the Operator Cues and Components Credited with Providing Operator Cues for Post Fire Safe Shutdown for each HFE.
Additionally, a table with cues associated with ANO-1 Fire HFEs is provided and documented in the ANO-1 Fire HRA Calculator file.
The ANO-1 HRA methodology was revised consistent with the approach used to address ANO-2 Include operator      One recovery action is incorporated into the PRA model for multiple      NFPA 805 LAR RAIs. The            This issue recovery actions that loss of DC breaker for 4160 bus events, which have an accident            revised methodology uses the      has no impact can restore the      sequence associated with them. The event found indicates that            NUREG-1921 methodology with        in STI change functions, systems,  recovery actions were incorporated for significant sequences rather      detailed Human Error Probabilities evaluations HRA-D1 or components on      than universally. The identification of all recovery actions used in the  (HEPs) developed for each          performed in an as-needed basis    model is documented in Attachment D of ANO-1 Fire HRA Notebook            Human Failure Event (HFE)          accordance provide a more        (Report 0247060006.03-U1) Most fire-specific recoveries used              credited in the FPRA. All events  with the realistic evaluation. screening values so this was set as CC-1.                                used in the FPRA have been        SFCP.
developed and documented using the same methods used for the internal events HEPs.
to 1CAN031801 Page 101 of 102 3.4    Identification of Key Assumptions The Initiative 5b is a risk-informed process that uses PRA model results to support a proposed STI change. The IDP uses the PRA results as an input to decide whether an STI change is warranted. The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized in the determination of the STI extension impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the STI assessment.
Any additional sensitivity studies identified for specific STI changes are also required per NEI 04-10, Revision 1. Therefore, results of the standby failure rate sensitivity study plus the results of any additional sensitivity studies identified during the performance of the reviews of gaps and open items as summarized in Sections 3.2 and 3.3 herein, will be documented and included in the results of the risk analysis submitted to the IDP.
3.5    External Events and Shutdown Considerations The NEI 04-10 methodology allows for STI change evaluations to be performed in the absence of quantifiable PRA models for all external hazards and shutdown. For those cases where the STI cannot be modeled in the plant PRA, or where a particular PRA model does not exist for a given hazard group, a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.
External hazards were evaluated in the ANO-1 Individual Plant Examination of External Events (IPEEE) submittal in response to the NRC IPEEE Program (Reference 15). The IPEEE Program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks. ANO-1 does not have a PRA model or applications associated with external hazards such as seismic, high wind, or external flooding, and quantitative results cannot be provided to support this STI effort. Therefore, a qualitative or bounding approach will be used to assess external event hazard risk at ANO-1 for STI changes.
Because ANO-1 does not have external hazards or shutdown PRA models, external hazards and shutdown screening evaluations are expected to be performed for STI changes in accordance with the guidance of NEI 04-10, Revision 1.
The ANO-1 shutdown safety program developed to support implementation of NUMARC 91-06 (Reference 16) is used for the shutdown risk evaluation, or an application-specific shutdown analysis may be performed for STI changes in accordance with the guidance of NEI 04-10, Revision 1. The ANO-1 shutdown safety program includes input from a Defense-in-Depth shutdown EOOS PRA model.
: 4. CONCLUSIONS The information presented herein demonstrate that the ANO-1 PRA technical adequacy and capability evaluations, as well as the maintenance and update processes conform to the ASME/ANS PRA Standard, which satisfies the guidance of RG 1.200, Revision 2.
to 1CAN031801 Page 102 of 102
: 5. REFERENCES
: 1. TSTF-425, Technical Specification Task Force - Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b, Revision 3, March 2009.
: 2. NEI 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1, April 2007.
: 3. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009.
: 4. ASME RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2009.
: 5. RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, May 2011.
: 6. 38-1290272-00, Arkansas Nuclear One - 1 Probabilistic Risk Assessment Peer Review Report, October 2002.
: 7. ANO-1 RG 1.200 PRA Peer Review Report Using ASME PRA Standard Requirements, August 2009.
: 8. ENTGANO150-REPT-001, Arkansas Nuclear One Unit 1 Internal Flooding Probabilistic Risk Assessment Peer Review, Revision 0, April 2017.
: 9. NUREG/CR-6850 - EPRI-1011089, Fire PRA Methodology for Nuclear Power Facilities, August 2005.
: 10. PRA-A1-05-005, ANO-1 Fire PRA Summary Report, Revision 1, 2016.
: 11. LTR-RAM-II-10-003, Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Arkansas Nuclear One Unit 1 Fire Probabilistic Risk Assessment, January 2010.
: 12. 0021-0022-005, Focused Peer Review of the Arkansas Nuclear One Unit 1 Fire Probabilistic Risk Assessment, Kleinsorg Group, May 2012.
: 13. 5384.R01.121129, Focused Scope Peer Review ANO-1 Fire PRA FSS-A, C, D, E and H, Kazarians & Associates, Inc., November 2012.
: 14. Arkansas Nuclear One Unit 1 Fire HRA Peer Review Report, Curtiss-Wright, June 2014.
: 15. Generic Letter 88-20, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4, June 1991.
: 16. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, December 1991.
: 17. 1CAN011401, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
Arkansas Nuclear One - Unit 1, January 29, 2014.
 
ATTACHMENT 3 1CAN031801 PROPOSED TECHNICAL SPECIFICATION CHANGES
 
Definitions 1.1 1.1 Definition (continued)
SHUTDOWN MARGIN (SDM)      SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
: a. All full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL RODS must be accounted for in the determination of SDM;
: b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level; and
: c. There is no change in APSR position.
STAGGERED TEST BASIS      A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER              THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
ANO-1                            1.1-5                      Amendment No. 215,218,
 
SDM 3.1.1 3.1  REACTIVITY CONTROL SYSTEMS 3.1.1  SHUTDOWN MARGIN (SDM)
LCO 3.1.1          The SDM shall be within the limit specified in the COLR.
APPLICABILITY:      MODES 3, 4, and 5.
ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME A. SDM not within limit.      A.1  Initiate boration to restore    15 minutes SDM to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.1.1.1        Verify SDM greater than or equal to the limit            24 hoursIn specified in the COLR.                                  accordance with the Surveillance Frequency Control Program ANO-1                                      3.1.1-1                        Amendment No. 215,
 
Reactivity Balance 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.1.2.1  ---------------------------------NOTES----------------------------
: 1. The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.
: 2.      This Surveillance is not required to be performed prior to entry into MODE 2.
Verify measured core reactivity balance is within                      Once prior to
            +/- 1% k/k of predicted values.                                          entering MODE 1 after each fuel loading AND
                                                                                    --------NOTE--------
Only required after 60 EFPD In accordance with the Surveillance Frequency Control Program31 EFPD thereafter ANO-1                                          3.1.2-2                              Amendment No. 215,
 
CONTROL ROD Group Alignment Limits 3.1.4 CONDITION                      REQUIRED ACTION                      COMPLETION TIME A. (continued)                  A.2.2.3 --------------NOTE-------------
Only required when THERMAL POWER is
                                        > 20% RTP.
Perform SR 3.2.5.1.                  72 hours B. Required Action and          B.1    Be in MODE 3.                      6 hours associated Completion Time for Condition A not met.
C. More than one CONTROL        C.1.1  Verify SDM to be within the        1 hour ROD inoperable, or not              limit provided in the COLR.
aligned within 6.5% of its group average height, or        OR both.
C.1.2  Initiate boration to restore        1 hour SDM to within limit.
AND C.2    Be in MODE 3.                      6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                          FREQUENCY SR 3.1.4.1      Verify individual CONTROL ROD positions are within            In accordance with 6.5% of their group average height.                          the Surveillance Frequency Control Program12 hours SR 3.1.4.2      Verify CONTROL ROD freedom of movement for                    In accordance with each individual CONTROL ROD that is not fully                the Surveillance inserted.                                                    Frequency Control Program92 days ANO-1                                      3.1.4-2                            Amendment No. 215,
 
Safety Rod Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR 3.1.5.1  Verify each safety rod is fully withdrawn.      In accordance with the Surveillance Frequency Control Program12 hours ANO-1                                    3.1.5-2              Amendment No. 215,
 
APSR Alignment Limits 3.1.6 3.1  REACTIVITY CONTROL SYSTEMS 3.1.6  AXIAL POWER SHAPING ROD (APSR) Alignment Limits LCO 3.1.6            Each APSR shall be OPERABLE and aligned to within 6.5% of its group average height.
APPLICABILITY:        MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION              COMPLETION TIME A. One APSR inoperable, or          A.1    Perform SR 3.2.5.1.      2 hours not aligned to within 6.5%
of its group average height,                                      AND or both.
2 hours after each APSR movement B. Require Action and                B.1    Be in MODE 3              6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.1.6.1          Verify position of each APSR is within 6.5% of the    In accordance with group average height.                                the Surveillance Frequency Control Program12 hours ANO-1                                          3.1.6-1                    Amendment No. 215,
 
Position Indicator Channels 3.1.7 3.1    REACTIVITY CONTROL SYSTEMS 3.1.7    Position Indicator Channels LCO 3.1.7                  One position indicator channel for each CONTROL ROD and APSR shall be OPERABLE.
APPLICABILITY:              MODES 1 and 2.
ACTIONS
--------------------------------------------------------------NOTES-------------------------------------------------------
Separate Condition entry is allowed for each CONTROL ROD and APSR.
CONDITION                                REQUIRED ACTION                            COMPLETION TIME A. The required position                  A.1        Declare the rod(s)                      Immediately indicator channel                                inoperable.
inoperable for one or more rods.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.1.7.1              Perform CHANNEL CHECK of required position                                In accordance with indicator channel.                                                        the Surveillance Frequency Control Program12 hours SR 3.1.7.2              Perform CHANNEL CALIBRATION of required                                  In accordance with position indicator channel.                                              the Surveillance Frequency Control Program18 months ANO-1                                                      3.1.7-1                                Amendment No. 215,
 
PHYSICS TESTS Exceptions - MODE 1 3.1.8 CONDITION                                REQUIRED ACTION                          COMPLETION TIME B. THERMAL POWER                        B.1      Suspend PHYSICS TESTS                  1 hour
                            > 85% RTP.                                      exceptions.
OR Nuclear overpower trip setpoint > 10% higher than PHYSICS TESTS power level.
OR Nuclear overpower trip setpoint > 90% RTP.
OR
                            ------------NOTE-------------
Only required when THERMAL POWER is
                            > 20% RTP.
LHR not within limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.1.8.1          Verify THERMAL POWER is  85% RTP.                                    In accordance with the Surveillance Frequency Control Program1 hour SR 3.1.8.2          ------------------------------NOTE----------------------------
Only required when THERMAL POWER is
                                              > 20% RTP.
Moved to Page 3.1.8-3 Perform SR 3.2.5.1.                                                  In accordance with the Surveillance Frequency Control Program2 hours SR 3.1.8.3          Verify nuclear overpower trip setpoint  10% RTP                      Within 8 hours prior higher than the THERMAL POWER at which the                            to performance of test is performed, with a maximum setting of                          PHYSICS TESTS at 90% RTP.                                                              each test plateau ANO-1                                                  3.1.8-2                                Amendment No. 215,
 
PHYSICS TESTS Exceptions - MODE 1 3.1.8 SURVEILLANCE                                                  FREQUENCY SR 3.1.8.2  ------------------------------NOTE-------------------------------
Only required when THERMAL POWER is
                                        > 20% RTP.
Moved from Page 3.1.8-2 Perform SR 3.2.5.1.                                                    In accordance with the Surveillance Frequency Control Program2 hours SR 3.1.8.3  Verify nuclear overpower trip setpoint  10% RTP                        Within 8 hours higher than the THERMAL POWER at which the test                        prior to is performed, with a maximum setting of 90% RTP.                        performance of PHYSICS TESTS at each test plateau SR 3.1.8.4  Verify SDM to be within the limits provided in the                      In accordance with COLR.                                                                  the Surveillance Frequency Control Program24 hours ANO-1                                          3.1.8-3                                Amendment No. 215,
 
PHYSICS TESTS Exceptions - MODE 2 3.1.9 CONDITION                        REQUIRED ACTION              COMPLETION TIME C. Nuclear overpower trip        C.1    Suspend PHYSICS TESTS        1 hour setpoint is not within limit.        exceptions.
OR Nuclear instrumentation high startup rate CONTROL ROD withdrawal inhibit inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.1.9.1        Verify THERMAL POWER is  5% RTP.                      In accordance with the Surveillance Frequency Control Program1 hour SR 3.1.9.2        Verify nuclear overpower trip setpoint is  5% RTP. Within 8 hours prior to performance of PHYSICS TESTS SR 3.1.9.3        Verify SDM to be within the limit provided in the      In accordance with COLR.                                                  the Surveillance Frequency Control Program24 hours ANO-1                                      3.1.9-2                      Amendment No. 215,
 
Regulating Rod Insertion Limits 3.2.1 CONDITION                      REQUIRED ACTION                  COMPLETION TIME D. Regulating rod groups          D.1    Initiate boration to restore    15 minutes inserted in unacceptable              SDM to within the limit operation region.                    provided in the COLR.
AND D.2.1 Restore regulating rod          2 hours groups to within restricted operation region.
OR D.2.2 Reduce THERMAL                  2 hours POWER to less than or equal to the THERMAL POWER allowed by the regulating rod group insertion limits.
E. Required Actions and          E.1    Be in MODE 3.                  6 hours associated Completion Times of Conditions C or D not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.2.1.1      Verify regulating rod groups are within the sequence      In accordance with and overlap limits as specified in the COLR.              the Surveillance Frequency Control Program12 hours SR 3.2.1.2      Verify regulating rod groups meet the insertion limits    In accordance with as specified in the COLR.                                  the Surveillance Frequency Control Program12 hours SR 3.2.1.3      Verify SDM  1% k/k.                                      Within 4 hours prior to achieving criticality ANO-1                                        3.2.1-2                        Amendment No. 215,
 
APSR Insertion Limits 3.2.2 3.2  POWER DISTRIBUTION LIMITS 3.2.2  AXIAL POWER SHAPING ROD (APSR) Insertion Limits LCO 3.2.2            APSRs shall be positioned within the limits specified in the COLR.
APPLICABILITY:      MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                      COMPLETION TIME A. APSRs not within limits. A.1    -------------NOTE---------------
Only required when THERMAL POWER is
                                        > 20% RTP.
Perform SR 3.2.5.1.                  Once per 2 hours AND A.2    Restore APSRs to within              24 hours limits.
B. Required Action and          B.1    Be in MODE 3.                        6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                            FREQUENCY SR 3.2.2.1        Verify APSRs are within acceptable limits specified in      In accordance with the COLR.                                                    the Surveillance Frequency Control Program12 hours ANO-1                                        3.2.2-1                            Amendment No. 215,
 
AXIAL POWER IMBALANCE Operating Limits 3.2.3 3.2  POWER DISTRIBUTION LIMITS 3.2.3    AXIAL POWER IMBALANCE Operating Limits LCO 3.2.3            AXIAL POWER IMBALANCE shall be maintained within the limits specified in the COLR.
APPLICABILITY:      MODE 1 with THERMAL POWER > 40% RTP.
ACTIONS CONDITION                    REQUIRED ACTION            COMPLETION TIME A. AXIAL POWER                    A.1  Perform SR 3.2.5.1.        Once per 2 hours IMBALANCE not within limits.                      AND A.2  Reduce AXIAL POWER          24 hours IMBALANCE to within limits.
B. Required Action and            B.1  Reduce THERMAL POWER        4 hours associated Completion              to  40% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.2.3.1        Verify AXIAL POWER IMBALANCE is within limits as    In accordance with specified in the COLR.                              the Surveillance Frequency Control Program12 hours ANO-1                                        3.2.3-1                  Amendment No. 215,
 
QPT 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.2.4.1  Verify QPT is within limits as specified in the COLR. In accordance with the Surveillance Frequency Control Program7 days AND When QPT has been restored to less than or equal to the steady state limit, 1 hour for 12 consecutive hours, or until verified acceptable at  95% RTP ANO-1                                    3.2.4-3                    Amendment No. 215,
 
RPS Instrumentation 3.3.1 CONDITION                                REQUIRED ACTION                            COMPLETION TIME D. As required by Required                D.1      Be in MODE 3.                            6 hours Action C.1 and referenced in                        AND Table 3.3.1-1.
D.2      Open all control rod drive                6 hours (CRD) trip breakers.
E. As required by Required                E.1      Open all CRD trip                        6 hours Action C.1 and                                  breakers.
referenced in Table 3.3.1-1.
F. As required by Required                F.1      Reduce THERMAL POWER                      6 hours Action C.1 and                                  < 45% RTP.
referenced in Table 3.3.1-1.
G. As required by Required                G.1      Reduce THERMAL POWER                      6 hours Action C.1 and                                  < 10% RTP.
referenced in Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS
-------------------------------------------------------------NOTE--------------------------------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply to each RPS Function.
SURVEILLANCE                                                      FREQUENCY SR 3.3.1.1              Perform CHANNEL CHECK.                                                In accordance with the Surveillance Frequency Control Program12 hours ANO-1                                                      3.3.1-2                                Amendment No. 215,
 
RPS Instrumentation 3.3.1 SURVEILLANCE                                                  FREQUENCY SR 3.3.1.2 -----------------------------NOTES---------------------------
: 1. Adjust power range channel output if the absolute difference is > 2% RTP.
: 2. Not required to be performed until 24 hours after THERMAL POWER is  20% RTP.
Compare results of calorimetric heat balance                        In accordance with calculation to power range channel output.                          the Surveillance Frequency Control Program96 hours AND Once within 24 hours after a THERMAL POWER change of 10% RTP SR 3.3.1.3 ------------------------------NOTES--------------------------
: 1. Adjust the power range channel imbalance output if the absolute value of the imbalance error is  2% RTP.
: 2. Not required to be performed until 24 hours after THERMAL POWER is  20% RTP.
Compare results of out of core measured AXIAL                      In accordance with POWER IMBALANCE to incore measured AXIAL                            the Surveillance POWER IMBALANCE.                                                    Frequency Control Program31 days SR 3.3.1.4 Perform CHANNEL FUNCTIONAL TEST.                                    In accordance with the Surveillance Frequency Control Program31 days ANO-1                                        3.3.1-3                            Amendment No. 215,
 
RPS Instrumentation 3.3.1 SURVEILLANCE                                                  FREQUENCY SR 3.3.1.5  ------------------------------NOTE----------------------------
Moved to Page 3.3.1-4 Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program18 months ANO-1                                          3.3.1-3                              Amendment No. 215,
 
RPS Instrumentation 3.3.1 SURVEILLANCE                                                  FREQUENCY Moved from Page 3.3.1-3 SR 3.3.1.5  ------------------------------NOTE-------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                            In accordance with the Surveillance Frequency Control Program18 months Table 3.3.1-1 Moved to Next (new) Page ANO-1                                          3.3.1-4                                Amendment No. 215,
 
RPS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Protection System Instrumentation APPLICABLE        CONDITIONS MODES OR          REFERENCED OTHER              FROM SPECIFIED          REQUIRED            SURVEILLANCE                ALLOWABLE FUNCTION                  CONDITIONS          ACTION C.1          REQUIREMENTS                  VALUE
: 1. Nuclear Overpower -
: a. High Setpoint                          1,2(a),3(d)            D                SR 3.3.1.1              104.9% RTP SR 3.3.1.2 SR 3.3.1.4 SR 3.3.1.5
: b.      Low Setpoint 2(b),3(b)            E                SR 3.3.1.1                5% RTP 4(b),5(b)                              SR 3.3.1.4 SR 3.3.1.5
: 2. RCS High Outlet Temperature                    1,2                D                SR 3.3.1.1                  618 &deg;F SR 3.3.1.4 SR 3.3.1.5 Moved from Page 3.3.1-4 to new Page 3.3.1-5
: 3. RCS High Pressure                          1,2(a),3(d)            D                SR 3.3.1.1                2355 psig SR 3.3.1.4 SR 3.3.1.5
: 4. RCS Low Pressure                              1,2(a)              D                SR 3.3.1.1                1800 psig SR 3.3.1.4 SR 3.3.1.5
: 5. RCS Variable Low Pressure                    1,2(a)              D                SR 3.3.1.1            As specified in the SR 3.3.1.4                COLR SR 3.3.1.5
: 6. Reactor Building High Pressure              1,2,3(c)              D                SR 3.3.1.1                18.7 psia SR 3.3.1.4 SR 3.3.1.5
: 7. Reactor Coolant Pump to                      1,2(a)              D                SR 3.3.1.1          55% RTP with one Power                                                                              SR 3.3.1.4        pump operating in each SR 3.3.1.5                loop.
: 8. Nuclear Overpower RCS Flow                    1,2(a)              D                SR 3.3.1.1            As specified in the and Measured AXIAL                                                              SR 3.3.1.3                COLR POWER IMBALANCE                                                                SR 3.3.1.4 SR 3.3.1.5
: 9. Main Turbine Trip (Oil                      45% RTP              F                SR 3.3.1.1                40.5 psig Pressure)                                                                          SR 3.3.1.4 SR 3.3.1.5
: 10. Loss of Main Feedwater Pumps              10% RTP              G                SR 3.3.1.1                55.5 psig (Control Oil Pressure)                                                        SR 3.3.1.4 SR 3.3.1.5
: 11. Shutdown Bypass RCS High                    2(b),3(b)            E                SR 3.3.1.1                1720 psig Pressure                                4(b),5(b)                              SR 3.3.1.4 SR 3.3.1.5 (a)        When not in shutdown bypass operation.
(b)        During shutdown bypass operation with any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(c)        With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(d)        With any CRD trip breaker in the closed position, the CRD system capable of rod withdrawal, and not in shutdown bypass operation.
ANO-1                                                          3.3.1-5                                Amendment No. 215,
 
RPS - RTM 3.3.3 CONDITION                REQUIRED ACTION              COMPLETION TIME C. Two or more RTMs          C.1  Open all CRD trip breakers. 6 hours inoperable in MODE 4 or 5.
OR OR C.2  Remove power from all      6 hours Required Action and              CRD trip breakers.
associated Completion Time not met in MODE 4 or 5.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.3.3.1      Perform CHANNEL FUNCTIONAL TEST.                In accordance with the Surveillance Frequency Control Program92 days ANO-1                                3.3.3-2                    Amendment No. 215,
 
CRD Trip Devices 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                FREQUENCY SR 3.3.4.1  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program92 days ANO-1                            3.3.4-3      Amendment No. 215,
 
ESAS Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                FREQUENCY SR 3.3.5.1  Perform CHANNEL CHECK.          In accordance with the Surveillance Frequency Control Program12 hours SR 3.3.5.2  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program31 days SR 3.3.5.3  Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program18 months ANO-1                            3.3.5-2      Amendment No. 215,
 
ESAS Manual Initiation 3.3.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE                  FREQUENCY SR 3.3.6.1  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program18 months ANO-1                            3.3.6-2        Amendment No. 215,
 
ESAS Actuation Logic 3.3.7 3.3    INSTRUMENTATION 3.3.7    Engineered Safeguards Actuation System (ESAS) Actuation Logic LCO 3.3.7                  The ESAS digital actuation logic channels shall be OPERABLE.
APPLICABILITY:              MODES 1 and 2, MODES 3 and 4 when associated engineered safeguards equipment is required to be OPERABLE.
ACTIONS
----------------------------------------------------------NOTE-----------------------------------------------------------
Separate Condition entry is allowed for each digital actuation logic channel.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more digital                        A.1      Place associated                        1 hour actuation logic channels                          component(s) in inoperable.                                        engineered safeguards configuration.
OR A.2      Declare the associated                  1 hour component(s) inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.3.7.1              Perform digital actuation logic CHANNEL                                  In accordance with FUNCTIONAL TEST.                                                        the Surveillance Frequency Control Program31 days ANO-1                                                      3.3.7-1                                Amendment No. 215,
 
DG LOPS 3.3.8 3.3    INSTRUMENTATION 3.3.8    Diesel Generator (DG) Loss of Power Start (LOPS)
LCO 3.3.8                  Two loss of voltage Function relays and two degraded voltage Function relays DG LOPS instrumentation per DG shall be OPERABLE.
APPLICABILITY:              MODES 1, 2, 3, and 4.
ACTIONS
----------------------------------------------------------NOTE----------------------------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more Functions                      A.1      Restore relay(s) to                      1 hour with one or more relays for                        OPERABLE status.
one or more DGs inoperable.
B. Required Action and                        B.1      Declare affected DG(s)                  Immediately associated Completion                              inoperable.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.3.8.1              Perform CHANNEL CHECK.                                                  In accordance with the Surveillance Frequency Control Program7 days ANO-1                                                      3.3.8-1                                  Amendment No. 215,
 
DG LOPS 3.3.8 SURVEILLANCE                                                  FREQUENCY SR 3.3.8.2 ------------------------------NOTE------------------------------
When DG LOPS instrumentation is placed in an inoperable status solely for performance of this Surveillance, entry into associated Conditions and Required Actions may be delayed up to 4 hours for the loss of voltage Function, provided the one remaining relay monitoring the Function for the bus is OPERABLE.
Perform CHANNEL CALIBRATION with setpoint                            In accordance with Allowable Value as follows:                                          the Surveillance Frequency Control
: a.      Degraded voltage  423.2 V and  436.0 V                      Program18 months with a time delay of 8 seconds +/- 1 second; and
: b.      Loss of voltage  1600 V and  3000 V with a time delay of  0.30 seconds and 0.98 seconds.
ANO-1                                        3.3.8-2                            Amendment No. 215,
 
Source Range Neutron Flux 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.3.9.1  Perform CHANNEL CHECK.                                                In accordance with the Surveillance Frequency Control Program12 hours SR 3.3.9.2  --------------------------------NOTE-----------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program18 months ANO-1                                          3.3.9-2                              Amendment No. 215,
 
Intermediate Range Neutron Flux 3.3.10 3.3  INSTRUMENTATION 3.3.10  Intermediate Range Neutron Flux LCO 3.3.10          One intermediate range neutron flux channel shall be OPERABLE.
APPLICABILITY:      MODE 2 MODES 3, 4, and 5 with any control rod drive (CRD) trip breaker in the closed position and the CRD System capable of rod withdrawal.
ACTIONS CONDITION                        REQUIRED ACTION                      COMPLETION TIME A. Required channel              -------------------NOTE-------------------
inoperable.                  Plant temperature changes are allowed provided the temperature change is accounted for in the SDM calculations.
                                  ---------------------------------------------- Immediately A.1      Suspend operations involving positive reactivity changes.
AND                                            1 hour A.2      Open CRD trip breakers.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.3.10.1      Perform CHANNEL CHECK.                                        In accordance with the Surveillance Frequency Control Program12 hours SR 3.3.10.2      Perform CHANNEL FUNCTIONAL TEST.                              In accordance with the Surveillance Frequency Control Program31 days ANO-1                                        3.3.10-1                            Amendment No. 215,
 
Intermediate Range Neutron Flux 3.3.10 SURVEILLANCE                                                  FREQUENCY SR 3.3.10.3 ------------------------------NOTE------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program18 months ANO-1                                        3.3.10-2                            Amendment No. 215,
 
EFIC System Instrumentation 3.3.11 CONDITION                                REQUIRED ACTION                          COMPLETION TIME E. Required Action and                        E.1      Reduce THERMAL                            6 hours associated Completion Time                        POWER to  10% RTP.
not met for Function 1.a or 1.d.
F. Required Action and                        F.1      Be in MODE 3.                            6 hours associated Completion Time not met for                        AND Functions 1.c, 2, or 3.
F.2    Reduce steam generator                    12 hours pressure to < 750 psig.
SURVEILLANCE REQUIREMENTS
-----------------------------------------------------------NOTE----------------------------------------------------------
Refer to Table 3.3.11-1 to determine which SRs shall be performed for each EFIC Function.
SURVEILLANCE                                                      FREQUENCY SR 3.3.11.1            Perform CHANNEL CHECK.                                                    In accordance with the Surveillance Frequency Control Program12 hours SR 3.3.11.2            Perform CHANNEL FUNCTIONAL TEST.(Notes 1 & 2)                              In accordance with the Surveillance Frequency Control Program31 days SR 3.3.11.3            Perform CHANNEL CALIBRATION.(Notes 1 & 2)                                  In accordance with the Surveillance Frequency Control Program18 months ANO-1                                                    3.3.11-2                          Amendment No. 215,227,
 
EFIC System Instrumentation 3.3.11 The following notes apply only to the SG Level - Low function:
Moved to Page 3.3.11-3 Note 1: If the as-found channel setpoints are conservative with respect to the Allowable Value but outside their predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoints are not conservative with respect to the Allowable Value, the channel shall be declared inoperable.
Note 2: The instrument channel setpoint(s) shall be reset to a value that is equal to or more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The Limiting Trip Setpoint and the methodology used to determine the Limiting Trip Setpoint and the predefined as-found acceptance criteria band are specified in the Bases.
ANO-1                                      3.3.11-2                  Amendment No. 215,227,
 
EFIC System Instrumentation 3.3.11 SURVEILLANCE REQUIREMENTS (continued)
The following notes apply only to the SG Level - Low function:
Moved from Page 3.3.11-2 Note 1: If the as-found channel setpoints are conservative with respect to the Allowable Value but outside their predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoints are not conservative with respect to the Allowable Value, the channel shall be declared inoperable.
Note 2: The instrument channel setpoint(s) shall be reset to a value that is equal to or more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The Limiting Trip Setpoint and the methodology used to determine the Limiting Trip Setpoint and the predefined as-found acceptance criteria band are specified in the Bases.
ANO-1                                      3.3.11-3                  Amendment No. 215,227,
 
EFIC System Instrumentation 3.3.11 Table 3.3.11-1 Emergency Feedwater Initiation and Control System Instrumentation APPLICABLE MODES OR REQUIRED    SURVEILLANCE    ALLOWABLE FUNCTION                OTHER CHANNELS    REQUIREMENTS      VALUES SPECIFIED CONDITIONS
: 1. EFW Initiation
: a. Loss of MFW Pumps        10% RTP                4      SR 3.3.11.1        55.5 psig (Control Oil Pressure)                                    SR 3.3.11.2 SR 3.3.11.3
: b. SG Level - Low              1,2,3            4 per SG    SR 3.3.11.1    9.34 inches(c,d)
Moved from Page 3.3.11-3 to new Page 3.3.11-4 SR 3.3.11.2 SR 3.3.11.3
: c. SG Pressure - Low        1,2,3(a)          4 per SG    SR 3.3.11.1        584.2 psig SR 3.3.11.2 SR 3.3.11.3
: d. RCP Status                10% RTP                4      SR 3.3.11.1          NA SR 3.3.11.2
: 2. EFW Vector Valve Control
: a. SG Pressure - Low          1,2,3(a)          4 per SG    SR 3.3.11.1        584.2 psig SR 3.3.11.2 SR 3.3.11.3
: b. SG Differential            1,2,3(a)                4      SR 3.3.11.1        150 psid Pressure - High                                          SR 3.3.11.2 SR 3.3.11.3
: 3. Main Steam Line Isolation
: a. SG Pressure - Low        1,2,3(a)(b)        4 per SG    SR 3.3.11.1        584.2 psig SR 3.3.11.2 SR 3.3.11.3 (a) When SG pressure  750 psig.
(b) Except when all associated valves are closed and deactivated.
(c) The SG Level - Low Limiting Trip Setpoint in accordance with NRC letter dated September 7, 2005, Technical Specification For Addressing Issues Related To Setpoint Allowable Values, is  10.42 inches.
(d) Includes an actuation time delay of  10.4 seconds.
ANO-1                                        3.3.11-4              Amendment No. 215,227,
 
EFIC Manual Initiation 3.3.12 CONDITION                  REQUIRED ACTION          COMPLETION TIME D. Required Action and          D.1  Be in MODE 3.            6 hours associated Completion Time not met for EFW        AND Initiation Function.
D.2  Be in MODE 4.            12 hours E. Required Action and          E.1  Be in MODE 3.            6 hours associated Completion Time not met for Main        AND Steam Line Isolation Function.                    E.2.1 Reduce steam generator  12 hours pressure to < 750 psig.
OR E.2.2 Close and deactivate all 12 hours associated valves.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                            FREQUENCY SR 3.3.12.1      Perform CHANNEL FUNCTIONAL TEST.            In accordance with the Surveillance Frequency Control Program31 days ANO-1                                  3.3.12-2                  Amendment No. 215,
 
EFIC Logic 3.3.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE                FREQUENCY SR 3.3.13.1  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program31 days ANO-1                            3.3.13-2      Amendment No. 215,
 
EFIC Vector Logic 3.3.14 3.3  INSTRUMENTATION 3.3.14  Emergency Feedwater Initiation and Control (EFIC) Vector Logic.
LCO 3.3.14        Four channels of the EFIC vector logic shall be OPERABLE.
APPLICABILITY:    MODES 1 and 2, MODE 3 when steam generator pressure is  750 psig.
ACTIONS CONDITION                      REQUIRED ACTION              COMPLETION TIME A. One vector logic channel      A.1    Restore channel to              72 hours inoperable.                        OPERABLE status.
B. Required Action and          B.1    Be in MODE 3.                  6 hours associated Completion Time not met.                AND B.2    Reduce steam generator          12 hours pressure to < 750 psig.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.3.14.1      Perform a CHANNEL FUNCTIONAL TEST.                    In accordance with the Surveillance Frequency Control Program31 days ANO-1                                    3.3.14-1                      Amendment No. 215,
 
PAM Instrumentation 3.3.15 CONDITION                                  REQUIRED ACTION                        COMPLETION TIME E. As required by Required                    E.1      Be in MODE 3.                            6 hours Action D.1 and referenced in Table 3.3.15-1.                      AND E.2      Be in MODE 4.                            12 hours F. As required by Required                    F.1      Initiate action to prepare              Immediately Action D.1 and referenced                          and submit a Special in Table 3.3.15-1.                                Report.
SURVEILLANCE REQUIREMENTS
----------------------------------------------------------NOTE-----------------------------------------------------------
These SRs apply to each PAM instrumentation Function in Table 3.3.15-1.
SURVEILLANCE                                                    FREQUENCY SR 3.3.15.1            Perform CHANNEL CHECK for each required                                  In accordance with instrumentation channel that is normally energized.                      the Surveillance Frequency Control Program31 days SR 3.3.15.2            -------------------------------NOTE------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                            In accordance with the Surveillance Frequency Control Program18 months ANO-1                                                      3.3.15-2                          Amendment No. 215,222,
 
Control Room Isolation - High Radiation 3.3.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.3.16.1  Perform CHANNEL CHECK.                                                In accordance with the Surveillance Frequency Control Program12 hours SR 3.3.16.2  -------------------------------NOTE------------------------------
When the Control Room Isolation - High Radiation instrumentation is placed in an inoperable status solely for performance of this Surveillance, entry into associated Conditions and Required Actions may be delayed for up to 3 hours.
Perform CHANNEL FUNCTIONAL TEST.                                      In accordance with the Surveillance Frequency Control Program31 days SR 3.3.16.3  Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program18 months ANO-1                                          3.3.16-2                            Amendment No. 215,
 
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.1    RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1            RCS DNB parameters (loop pressure, hot leg temperature, and RCS total flow rate) shall be within the limits specified in the COLR.
APPLICABILITY:        MODE 1.
                      ------------------------------------------------NOTE-----------------------------------------
RCS loop pressure limit does not apply during pressure transients due to a THERMAL POWER change > 5% RTP per minute.
ACTIONS CONDITION                              REQUIRED ACTION                          COMPLETION TIME A. One or more RCS DNB                A.1      Restore RCS DNB                          2 hours parameters not within                        parameter(s) to within limits.                                      limit.
B. Required Action and                B.1      Be in MODE 2.                            6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.4.1.1        ---------------------------------NOTE----------------------------
With three RCPs operating, the limits are applied to the loop with two RCPs in operation.
Verify RCS loop pressure is within the limit specified                    In accordance with in the COLR.                                                              the Surveillance Frequency Control Program12 hours ANO-1                                                3.4.1-1                                Amendment No. 215,
 
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE                                                FREQUENCY SR 3.4.1.2 ---------------------------------NOTE---------------------------
With three RCPs operating, the limits are applied to the loop with two RCPs in operation.
Verify RCS hot leg temperature is within the limit                    In accordance with specified in the COLR.                                                the Surveillance Frequency Control Program12 hours SR 3.4.1.3 Verify RCS total flow is within the limit specified in                In accordance with the COLR.                                                            the Surveillance Frequency Control Program12 hours SR 3.4.1.4 ---------------------------------NOTE---------------------------
Only required to be performed when stable thermal conditions are established at  90% RTP.
Verify RCS total flow rate is within the limit specified              In accordance with in the COLR by measurement.                                          the Surveillance Frequency Control Program18 months ANO-1                                        3.4.1-2                            Amendment No. 215,
 
RCS Minimum Temperature for Criticality 3.4.2 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.2  RCS Minimum Temperature for Criticality LCO 3.4.2              The RCS average temperature (Tavg) shall be  525 &deg;F.
APPLICABILITY:          MODE 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                COMPLETION TIME A. Tavg not within limit.        A.1    Be in MODE 3.                30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.4.2.1          Verify RCS Tavg  525 &deg;F.                            In accordance with the Surveillance Frequency Control Program12 hours ANO-1                                        3.4.2-1                      Amendment No. 215,
 
RCS P/T Limits 3.4.3 CONDITION                                REQUIRED ACTION                        COMPLETION TIME D. -------------NOTE-------------          D.1      Initiate action to restore            Immediately Required Action D.2 shall                        parameter(s) to within limit.
be completed whenever this Condition is entered.              AND D.2      Determine RCS is                      Prior to entering Requirements of LCO not                          acceptable for continued              MODE 4 met in other than MODE 1,                        operation.
2, 3, or 4.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.3.1          ------------------------------NOTE-----------------------------
Only required to be performed during RCS heatup operations with fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                        In accordance with heatup rates are within the limits specified in                      the Surveillance Figure 3.4.3-1.                                                      Frequency Control Program30 minutes SR 3.4.3.2          ------------------------------NOTE-----------------------------
Only required to be performed during RCS cooldown operations with fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                        In accordance with cooldown rates are within the limits specified in                    the Surveillance Figure 3.4.3-2.                                                      Frequency Control Program30 minutes SR 3.4.3.3          ------------------------------NOTE-----------------------------
Moved to Page 3.4.3-3 Only required to be performed during RCS heatup and cooldown operations with no fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                        In accordance with cooldown rates are within the limits specified in                    the Surveillance Figure 3.4.3-3.                                                      Frequency Control Program30 minutes ANO-1                                                  3.4.3-2                              Amendment No. 215,
 
RCS P/T Limits 3.4.3 SURVEILLANCE                                                  FREQUENCY Moved from Page 3.4.3-2 SR 3.4.3.3  ------------------------------NOTE-----------------------------
Only required to be performed during RCS heatup and cooldown operations with no fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                        In accordance with cooldown rates are within the limits specified in                    the Surveillance Figure 3.4.3-3.                                                      Frequency Control Program30 minutes SR 3.4.3.4  ------------------------------NOTE----------------------------
Only required to be performed during PHYSICS TESTS with RCS temperature  525 &deg;F.
Verify RCS pressure and RCS temperature are                          In accordance with within the criticality limits specified in Figure 3.4.3-1.            the Surveillance Frequency Control Program30 minutes ANO-1                                          3.4.3-3                              Amendment No. 215,
 
RCS Loops - MODES 1 and 2 3.4.4 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.4  RCS Loops - MODES 1 and 2 LCO 3.4.4            Two RCS Loops shall be in operation, with:
: a. Four reactor coolant pumps (RCPs) operating; or
: b. Three RCPs operating and THERMAL POWER restricted as specified in the COLR.
APPLICABILITY:      MODES 1 and 2.
ACTIONS CONDITION                        REQUIRED ACTION            COMPLETION TIME A. One RCP not in operation      A.1      Restore one non-operating  18 hours in each loop.                          RCP to operation.
B. Required Action and          B.1      Be in MODE 3.              6 hours associated Completion Time of Condition A not met.
OR LCO not met for reasons other than Condition A.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.4.4.1        Verify required RCS loops are in operation.          In accordance with the Surveillance Frequency Control Program12 hours ANO-1                                          3.4.4-1                    Amendment No. 215,
 
RCS Loops - MODE 3 3.4.5 CONDITION                              REQUIRED ACTION                        COMPLETION TIME C. Two RCS loops inoperable. C.1              Suspend operations that                Immediately would cause introduction OR                                        into the RCS, coolant with boron concentration less Required RCS loop not in                  than required to meet SDM operation.                                of LCO 3.1.1.
AND C.2      Initiate action to restore one        Immediately RCS loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.5.1    Verify required RCS loop is in operation.                              In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.5.2    --------------------------------NOTE----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required pump.                                the Surveillance Frequency Control Program7 days ANO-1                                            3.4.5-2                              Amendment No. 215,
 
RCS Loops - MODE 4 3.4.6 CONDITION                              REQUIRED ACTION                      COMPLETION TIME B. Two required loops                B.1      Suspend operations that                Immediately inoperable.                                would cause introduction into the RCS, coolant with OR                                          boron concentration less than required to meet SDM Required loop not in                        of LCO 3.1.1.
operation.
AND B.2      Initiate action to restore one          Immediately loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.6.1      Verify required DHR or RCS loop is in operation.                      In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.6.2      -------------------------------NOTE-----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required pump.                                the Surveillance Frequency Control Program7 days ANO-1                                            3.4.6-2                                Amendment No. 215,
 
RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.7.1  Verify required DHR loop is in operation.                            In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.7.2  Verify required SG secondary side water levels are                    In accordance with 20 inches.                                                          the Surveillance Frequency Control Program12 hours SR 3.4.7.3  --------------------------------NOTE-----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required DHR pump.                            the Surveillance Frequency Control Program7 days ANO-1                                          3.4.7-3                            Amendment No. 215,
 
RCS Loops - MODE 5, Loops Not Filled 3.4.8 CONDITION                              REQUIRED ACTION                        COMPLETION TIME B. No required DHR loop              B.1      Suspend operations that              Immediately OPERABLE.                                    would cause introduction into the RCS, coolant with OR                                          boron concentration less than required to meet SDM Required DHR loop not in                    of LCO 3.1.1.
operation.
AND B.2      Suspend all operations                Immediately involving reduction in RCS water volume.
AND B.3      Initiate action to restore one        Immediately DHR loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.8.1      Verify required DHR loop is in operation.                            In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.8.2      --------------------------------NOTE-----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required DHR pump.                            the Surveillance Frequency Control Program7 days ANO-1                                              3.4.8-2                              Amendment No. 215,
 
Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.4.9.1  Verify pressurizer water level  320 inches. In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.9.2  Verify capacity of ES bus powered pressurizer  In accordance with heaters  126 kW.                              the Surveillance Frequency Control Program18 months ANO-1                                  3.4.9-2            Amendment No. 215,241,
 
LTOP System 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.4.11.1  Verify pressurizer level does not represent a water                  30 minutes during solid condition.                                                      RCS heatup and cooldown AND In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.11.2  Verify HPI is deactivated.                                            In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.11.3  Verify each pressurized CFT is isolated.                              In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.11.4  --------------------------------NOTE-----------------------------
Verification of locked, sealed, or otherwise secured open vent path(s) only required to be performed every 31 days.
Verify OPERABLE pressure relief capability.                          In accordance with the Surveillance Frequency Control Program12 hours SR 3.4.11.5  Perform CHANNEL CALIBRATION of ERV opening                            In accordance with circuitry.                                                            the Surveillance Frequency Control Program18 months ANO-1                                          3.4.11-3                            Amendment No. 215,
 
RCS Specific Activity 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.12.1  ----------------------------NOTE----------------------------------
Only required to be performed in MODE 1 and 2, MODE 3 with RCS average temperature  500 &deg;F.
Verify reactor coolant DOSE EQUIVALENT XE-133                          In accordance with specific activity  2200 &#xb5;Ci/gm.                                        the Surveillance Frequency Control Program7 days SR 3.4.12.2  Verify reactor coolant DOSE EQUIVALENT I-131                            In accordance with specific activity  1.0 &#xb5;Ci/gm.                                        the Surveillance Frequency Control Program14 days ANO-1                                          3.4.12-2                    Amendment No. 215,238,243,
 
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.13.1 ------------------------------NOTES-----------------------------
: 1. Not required to be performed until 12 hours after establishment of steady state operation at or near operating pressure.
: 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by                      In accordance with performance of an RCS water inventory balance.                          the Surveillance Frequency Control Program72 hours SR 3.4.13.2 ------------------------------NOTE-------------------------------
Not required to be performed until 12 hours after establishment of steady state operation.
Verify primary to secondary LEAKAGE is                                  In accordance with 150 gallons per day through any one SG.                              the Surveillance Frequency Control Program72 hours ANO-1                                          3.4.13-2                          Amendment No. 215,224,
 
RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.14.1  ----------------------------NOTE----------------------------------
Not required to be performed in MODES 3 and 4.
Verify leakage from each RCS pressure isolation                        In accordance with check valve, or pair of check valves, as applicable, is                the INSERVICE less than or equal to an equivalent of the Allowable                    TESTING Leakage Limit identified below at a differential test                  PROGRAM pressure  150 psid.
AND Pressure Isolation                      Allowable Check Valve(s)                    Leakage Limit                      Once prior to entering MODE 2 DH-14A                              5 gpm                          whenever the unit DH-13A and DH-17                    5 gpm total                    has been in DH-14B                              5 gpm                          MODE 5 for 7 days DH-13B and DH-18                    5 gpm total                    or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2  Verify DHR System autoclosure interlock prevents                        In accordance with the valves from being opened with a simulated or                        the Surveillance actual high RCS pressure signal.                                        Frequency Control Program18 months SR 3.4.14.3  Verify DHR System autoclosure interlock causes                          In accordance with the valves to close automatically with a simulated or                  the Surveillance actual high RCS pressure signal:                                        Frequency Control Program18 months
: a. 340 psig for one valve; and
: b. 400 psig for the other valve.
SR 3.4.14.4  Verify DHR System autoclosure interlock prevents                        In accordance with the valves from being opened with a simulated or                        the Surveillance actual Core Flood Tank isolation valve not closed                    Frequency Control signal.                                                                Program18 months SR 3.4.14.5  Verify DHR System autoclosure interlock causes the                      In accordance with valves to close automatically with a simulated or                      the Surveillance actual Core Flood Tank isolation valve not closed                    Frequency Control signal.                                                                Program18 months ANO-1                                          3.4.14-2                          Amendment No. 215,257,
 
RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.4.15.1  Perform CHANNEL CHECK of required reactor          In accordance with building atmosphere radioactivity monitor.        the Surveillance Frequency Control Program12 hours SR 3.4.15.2  Perform CHANNEL FUNCTIONAL TEST of                In accordance with required reactor building atmosphere radioactivity the Surveillance monitor.                                          Frequency Control Program92 days SR 3.4.15.3  Perform CHANNEL CALIBRATION of required            In accordance with reactor building atmosphere radioactivity monitor. the Surveillance Frequency Control Program 18 months SR 3.4.15.4  Perform CHANNEL CALIBRATION of required            In accordance with reactor building sump monitor.                    the Surveillance Frequency Control Program18 months ANO-1                                  3.4.15-3                  Amendment No. 215,
 
CFTs 3.5.1 3.5    EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1    Core Flood Tanks (CFTs)
LCO 3.5.1            Two CFTs shall be OPERABLE.
APPLICABILITY:        MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) pressure > 800 psig.
ACTIONS CONDITION                      REQUIRED ACTION          COMPLETION TIME A. One CFT inoperable due to A.1      Restore boron              72 hours boron concentration not            concentration to within within limits.                    limits.
B. One CFT inoperable for      B.1  Restore CFT to            1 hour reasons other than                OPERABLE status.
Condition A.
C. Required Action and          C.1  Be in MODE 3.              6 hours associated Completion Time of Condition A or B    AND not met.
C.2  Reduce RCS pressure to OR                                  800 psig.                12 hours Two CFTs inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.5.1.1        Verify each CFT isolation valve is fully open. In accordance with the Surveillance Frequency Control Program12 hours Moved to Page 3.5.1-2 SR 3.5.1.2        Verify borated water volume in each CFT is        In accordance with 970 ft3 and  1110 ft3.                        the Surveillance Frequency Control Program12 hours ANO-1                                        3.5.1-1                  Amendment No. 215,
 
CFTs 3.5.1 SURVEILLANCE                              FREQUENCY Moved to Page 3.5.1-2 SR 3.5.1.3  Verify nitrogen cover pressure in each CFT is  In accordance with 560 psig and  640 psig.                      the Surveillance Frequency Control Program12 hours ANO-1                                  3.5.1-1                Amendment No. 215,
 
CFTs 3.5.1 SURVEILLANCE                                FREQUENCY SR 3.5.1.2  Verify borated water volume in each CFT is        In accordance with Moved from Page 3.5.1-1 970 ft3 and  1110 ft3.                        the Surveillance Frequency Control Program12 hours SR 3.5.1.3  Verify nitrogen cover pressure in each CFT is    In accordance with 560 psig and  640 psig.                        the Surveillance Frequency Control Program12 hours SR 3.5.1.4  Verify boron concentration in each CFT is        In accordance with 2270 ppm.                                      the Surveillance Frequency Control Program31 days AND
                                                                                          ---------NOTE-------
Only required to be performed for affected CFT Once within 12 hours after each solution level increase of 0.2 feet that is not the result of addition from a borated water source of known concentration 2270 ppm SR 3.5.1.5  Verify power is removed from each CFT isolation  In accordance with valve operator.                                  the Surveillance Frequency Control Program31 days ANO-1                                  3.5.1-2                  Amendment No. 215,
 
ECCS - Operating 3.5.2 3.5    EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2    ECCS - Operating LCO 3.5.2            Two ECCS trains shall be OPERABLE.
APPLICABILITY:        MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) temperature > 350 &deg;F.
ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME A. One or more trains          A.1    Restore train(s) to              72 hours inoperable.                        OPERABLE status.
B. Required Action and          B.1    Be in MODE 3.                    6 hours associated Completion Time not met.                AND B.2    Reduce RCS temperature            12 hours to  350 &deg;F.
C. Less than 100% of the        C.1    Enter LCO 3.0.3.                  Immediately ECCS flow equivalent to a single OPERABLE train available.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.5.2.1        Verify each ECCS manual, power operated, and              In accordance with automatic valve in the flow path, that is not locked,    the Surveillance sealed, or otherwise secured in position, is in the      Frequency Control Moved to Page 3.5.1-2 correct position.                                        Program31 days SR 3.5.2.2        Verify each ECCS pump's developed head at the            In accordance with test flow point is greater than or equal to the          the INSERVICE required developed head.                                  TESTING PROGRAM ANO-1                                        3.5.2-1                      Amendment No. 215,257,
 
ECCS - Operating 3.5.2 Moved from Page 3.5.1-1 SURVEILLANCE                                      FREQUENCY SR 3.5.2.2  Verify each ECCS pump's developed head at the          In accordance with test flow point is greater than or equal to the        the INSERVICE required developed head.                                TESTING PROGRAM SR 3.5.2.3  Verify each ECCS automatic valve in the flow path      In accordance with that is not locked, sealed, or otherwise secured in    the Surveillance position, actuates to the correct position on an        Frequency Control actual or simulated actuation signal.                  Program18 months SR 3.5.2.4  Verify each ECCS pump starts automatically on an        In accordance with actual or simulated actuation signal.                  the Surveillance Frequency Control Program18 months SR 3.5.2.5  Verify, by visual inspection, each ECCS train          In accordance with reactor building sump suction inlet is not restricted  the Surveillance by debris and screens show no evidence of              Frequency Control structural distress or abnormal corrosion.              Program18 months ANO-1                                  3.5.2-2                        Amendment No. 215,
 
BWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.5.4.1  -------------------------------NOTE-----------------------------
Only required to be performed when ambient air temperature is < 40 &deg;F or > 110 &deg;F.
Verify BWST borated water temperature is                              In accordance with 40 &deg;F and  110 &deg;F.                                                  the Surveillance Frequency Control Program24 hours SR 3.5.4.2  Verify BWST borated water level is                                    In accordance with 38.4 feet and  42 feet.                                            the Surveillance Frequency Control Program7 days SR 3.5.4.3  Verify BWST boron concentration is                                    In accordance with 2270 ppm and  2670 ppm.                                            the Surveillance Frequency Control Program7 days ANO-1                                          3.5.4-2                            Amendment No. 215,253,
 
Reactor Building Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.6.2.1  -------------------------------NOTE-----------------------------
: 1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
: 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.
Perform required air lock leakage rate testing in                      In accordance with accordance with the Reactor Building Leakage Rate                      the Reactor Building Testing Program.                                                      Leakage Rate Testing Program SR 3.6.2.2  Verify only one door in the air lock can be opened at                  In accordance with a time.                                                                the Surveillance Frequency Control Program18 months ANO-1                                          3.6.2-4                              Amendment No. 215,
 
Reactor Building Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.6.3.1  Verify each reactor building purge isolation valve is                In accordance with closed.                                                              the Surveillance Frequency Control Program31 days SR 3.6.3.2  -------------------------------NOTE----------------------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each reactor building isolation manual valve                  In accordance with and blind flange that is located outside the reactor                the Surveillance building and not locked, sealed, or otherwise                        Frequency Control secured, and is required to be closed during                        Program31 days accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.
SR 3.6.3.3  -------------------------------NOTE----------------------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each reactor building isolation manual valve                  Prior to entering and blind flange that is located inside the reactor                  MODE 4 from building and not locked, sealed, or otherwise                        MODE 5 if not secured, and required to be closed during accident                  performed within conditions is closed, except for reactor building                    the previous isolation valves that are open under administrative                  92 days controls.
SR 3.6.3.4  Verify the isolation time of each automatic power                    In accordance with operated reactor building isolation valve is within                  the INSERVICE limits.                                                              TESTING PROGRAM SR 3.6.3.5  Verify each automatic reactor building isolation                    In accordance with valve that is not locked, sealed, or otherwise                      the Surveillance secured in position, actuates to the isolation                      Frequency Control position on an actual or simulated actuation signal.                Program18 months ANO-1                                          3.6.3-4                      Amendment No. 215,253,257,
 
Reactor Building Pressure 3.6.4 3.6  REACTOR BUILDING SYSTEMS 3.6.4  Reactor Building Pressure LCO 3.6.4              Reactor building pressure shall be  -1.0 psig and  +3.0 psig.
APPLICABILITY:          MODES 1, 2, 3, and 4.
ACTIONS CONDITION                          REQUIRED ACTION                        COMPLETION TIME A. Reactor building pressure      A.1  Restore reactor building                1 hour not within limits.                    pressure to within limits.
B. Required Action and            B.1  Be in MODE 3.                            6 hours associated Completion Time not met.                  AND B.2  ---------------NOTE--------------
LCO 3.0.4.a is not applicable when entering Mode 4.
Be in MODE 4.                            12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.6.4.1          Verify reactor building pressure is  -1.0 psig and              In accordance with
                      +3.0 psig.                                                    the Surveillance Frequency Control Program12 hours ANO-1                                          3.6.4-1                            Amendment No. 215,253,
 
Reactor Building Spray and Cooling System 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.6.5.1  Verify each reactor building spray manual, power          In accordance with operated, and automatic valve in each required flow        the Surveillance path that is not locked, sealed, or otherwise secured      Frequency Control in position is in the correct position.                    Program31 days SR 3.6.5.2  Operate each required reactor building cooling train      In accordance with fan unit for  15 minutes.                                the Surveillance Frequency Control Program31 days SR 3.6.5.3  Verify each required reactor building cooling train        In accordance with cooling water flow rate is  1200 gpm.                    the Surveillance Frequency Control Program31 days SR 3.6.5.4  Verify each required reactor building spray pump's        In accordance with developed head at the flow test point is greater than      the INSERVICE or equal to the required developed head.                  TESTING PROGRAM SR 3.6.5.5  Verify each automatic reactor building spray valve in      In accordance with each required flow path that is not locked, sealed, or    the Surveillance otherwise secured in position, actuates to the correct    Frequency Control position on an actual or simulated actuation signal.      Program18 months SR 3.6.5.6  Verify each required reactor building spray pump          In accordance with starts automatically on an actual or simulated            the Surveillance actuation signal.                                          Frequency Control Program18 months SR 3.6.5.7  Verify each required reactor building cooling train        In accordance with starts automatically on an actual or simulated            the Surveillance actuation signal.                                          Frequency Control Program18 months SR 3.6.5.8  Verify each spray nozzle is unobstructed.                  Following maintenance which could result in nozzle blockage ANO-1                                    3.6.5-3              Amendment No. 215,233,257,
 
Spray Additive System 3.6.6 3.6    REACTOR BUILDING SYSTEMS 3.6.6    Spray Additive System LCO 3.6.6            The Spray Additive System shall be OPERABLE.
APPLICABILITY:      MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                    COMPLETION TIME A. Spray Additive System      A.1  Restore Spray Additive              72 hours inoperable.                      System to OPERABLE status.
B. Required Action and        B.1  Be in MODE 3.                      6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                          FREQUENCY SR 3.6.6.1        Verify each Spray Additive System manual, power            In accordance with operated, and automatic valve in the flow path that is      the Surveillance not locked, sealed, or otherwise secured in position is    Frequency Control in the correct position.                                    Program31 days SR 3.6.6.2        Verify sodium hydroxide tank solution volume is            In accordance with 9000 gallons.                                            the Surveillance Frequency Control Program184 days Moved to new Page 3.6.6-2 SR 3.6.6.3        Verify sodium hydroxide tank solution concentration is      In accordance with
                                              > 6.0 wt% and < 8.5 wt.% NaOH.                              the Surveillance Frequency Control Program184 days SR 3.6.6.4        Verify each Spray Additive System automatic valve in        In accordance with the flow path actuates to the correct position on an        the Surveillance actual or simulated actuation signal.                      Frequency Control Program18 months ANO-1                                        3.6.6-1                  Amendment No. 215,234,
 
Spray Additive System 3.6.6 SURVEILLANCE                                        FREQUENCY SR 3.6.6.3  Verify sodium hydroxide tank solution concentration is    In accordance with Moved from Page 3.6.6-1
                                        > 6.0 wt% and < 8.5 wt.% NaOH.                            the Surveillance Frequency Control Program184 days SR 3.6.6.4  Verify each Spray Additive System automatic valve in      In accordance with the flow path actuates to the correct position on an      the Surveillance actual or simulated actuation signal.                      Frequency Control Program18 months ANO-1                                  3.6.6-2                  Amendment No. 215,234,
 
MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.2.1  --------------------------------NOTE-----------------------------
Only required to be performed in MODES 1 and 2.
Verify isolation time of each MSIV is within the limits                In accordance with specified in the INSERVICE TESTING PROGRAM.                            the INSERVICE TESTING PROGRAM SR 3.7.2.2  --------------------------------NOTE-----------------------------
: 1. Only required to be performed in MODES 1 and 2.
: 2. Not required to be met when SG pressure is
                    < 750 psig.
Verify each MSIV actuates to the isolation position                    In accordance with on an actual or simulated actuation signal.                            the Surveillance Frequency Control Program18 months ANO-1                                          3.7.2-2                          Amendment No. 215,257,
 
MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves 3.7.3 SURVEILLANCE                                                FREQUENCY SR 3.7.3.2 --------------------------------NOTES---------------------------
: 1. Only required to be performed in MODES 1 and 2.
: 2. Not required to be met when SG pressure is
                  < 750 psig.
Verify that each MFIV, Main Feedwater Block Valve,                    In accordance with Low Load Feedwater Control Valve and Startup                          the Surveillance Feedwater Control Valve actuates to the isolation                    Frequency Control position on an actual or simulated actuation signal.                  Program18 months ANO-1                                        3.7.3-3                            Amendment No. 215,
 
Secondary Specific Activity 3.7.4 3.7  PLANT SYSTEMS 3.7.4    Secondary Specific Activity LCO 3.7.4              The specific activity of the secondary coolant shall be  0.1 Ci/gm DOSE EQUIVALENT I-131.
APPLICABILITY:          MODES 1, 2, 3, and 4.
ACTIONS CONDITION                          REQUIRED ACTION                COMPLETION TIME A. Specific activity not within    A.1    Be in MODE 3.                  6 hours limit.
AND A.2    Be in MODE 5.                  36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.7.4.1          Verify the specific activity of the secondary coolant    In accordance with is  0.1 Ci/gm DOSE EQUIVALENT I-131.                    the Surveillance Frequency Control Program31 days ANO-1                                            3.7.4-1                  Amendment No. 215,238,
 
EFW System 3.7.5 CONDITION                              REQUIRED ACTION                      COMPLETION TIME C. Required Action and                C.1      Be in MODE 3.                          6 hours associated Completion Time of Condition A or B          AND not met.
C.2      Be in MODE 4.                          18 hours D. Two EFW trains                    D.1      -------------NOTE----------------
inoperable in MODE 1, 2,                    LCO 3.0.3 and all other or 3.                                      LCO Required Actions requiring MODE changes are suspended until one EFW train is restored to OPERABLE status.
Initiate action to restore              Immediately one EFW train to OPERABLE status.
E. Required EFW train                E.1      Initiate action to restore              Immediately inoperable in MODE 4.                      EFW train to OPERABLE status.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.5.1      Verify each EFW manual, power operated, and                            In accordance with automatic valve in each water flow path and in both                    the Surveillance steam supply flow paths to the steam turbine driven                    Frequency Control pump, that is not locked, sealed, or otherwise                        Program31 days secured in position, is in the correct position.
SR 3.7.5.2      -------------------------------NOTE------------------------------
Not required to be performed for the turbine driven EFW pump, until 24 hours after reaching  750 psig in the steam generators.
Verify the developed head of each EFW pump at                          In accordance with the flow test point is greater than or equal to the                    the INSERVICE required developed head.                                              TESTING PROGRAM ANO-1                                              3.7.5-2                          Amendment No. 215,257,
 
EFW System 3.7.5 SURVEILLANCE                                                FREQUENCY SR 3.7.5.3 -------------------------------NOTE------------------------------
Not required to be met in MODE 4 when steam generator is relied upon for heat removal.
Verify each EFW automatic valve that is not locked,                  In accordance with sealed, or otherwise secured in position, actuates to                the Surveillance the correct position on an actual or simulated                        Frequency Control actuation signal.                                                    Program18 months SR 3.7.5.4 -------------------------------NOTE------------------------------
Not required to be met in MODE 4 when steam generator is relied upon for heat removal.
Verify each EFW pump starts automatically on an                      In accordance with actual or simulated actuation signal.                                the Surveillance Frequency Control Program18 months SR 3.7.5.5 Verify proper alignment of the required EFW flow                      Prior to entering paths by verifying manual valve alignment from the                    MODE 2 whenever Q condensate storage tank to each steam                            the unit has been in generator.                                                            MODE 5, MODE 6, or defueled for a cumulative period of > 30 days SR 3.7.5.6 Verify that feedwater is delivered to each steam                      In accordance with generator using the motor-driven EFW pump.                            the Surveillance Frequency Control Program18 months ANO-1                                        3.7.5-3                            Amendment No. 215,
 
QCST 3.7.6 3.7  PLANT SYSTEMS 3.7.6  Q Condensate Storage Tank (QCST)
LCO 3.7.6            The QCST shall be OPERABLE.
APPLICABILITY:      MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
ACTIONS CONDITION                        REQUIRED ACTION            COMPLETION TIME A. The QCST inoperable.          A.1    Verify by administrative    4 hours means OPERABILITY of backup water supply.        AND Once per 12 hours thereafter AND A.2    Restore QCST to            7 days OPERABLE status.
B. Required Action and          B.1    Be in MODE 3.              6 hours associated Completion Time not met.                AND B.2    Be in MODE 4 without        24 hours reliance on steam generator for heat removal.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.7.6.1        Verify QCST volume is  267,000 gallons when        In accordance with required for both units and  107,000 gallons when  the Surveillance only required for Unit 1.                            Frequency Control Program12 hours ANO-1                                        3.7.6-1                    Amendment No. 215,
 
SWS 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.7.1  -------------------------------NOTE-----------------------------
Isolation of SWS flow to individual components does not render the SWS inoperable.
Verify each SWS manual, power operated, and                            In accordance with automatic valve in the flow path servicing safety                      the Surveillance related equipment, that is not locked, sealed, or                      Frequency Control otherwise secured in position, is in the correct                      Program31 days position.
SR 3.7.7.2  Verify each SWS automatic valve in the flow path                      In accordance with that is not locked, sealed, or otherwise secured in                    the Surveillance position, actuates to the correct position on an                      Frequency Control actual or simulated actuation signal.                                  Program18 months SR 3.7.7.3  Verify each required SWS pump starts                                  In accordance with automatically on an actual or simulated signal.                        the Surveillance Frequency Control Program18 months ANO-1                                          3.7.7-2                            Amendment No. 215,218,
 
ECP 3.7.8 3.7    PLANT SYSTEMS 3.7.8    Emergency Cooling Pond (ECP)
LCO 3.7.8            The ECP shall be OPERABLE.
APPLICABILITY:        MODES 1, 2, 3, and 4.
ACTIONS CONDITION                            REQUIRED ACTION                          COMPLETION TIME A. Degradation of the ECP            A.1    Determine ECP remains                    7 days noted pursuant to                        acceptable for continued SR 3.7.8.4 below or by                    operation.
other inspection.
B. Required Action and              B.1    Be in MODE 3.                            6 hours associated Completion Time of Condition A not          AND met.
B.2    Be in MODE 5.                            36 hours OR LCO not met for reasons other than Condition A.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.8.1        Verify that the indicated water level of the ECP is                    In accordance with greater than or equal to that required for an ECP                      the Surveillance volume of 70 acre-ft.                                                  Frequency Control Program24 hours Moved to new Page 3.7.8-2 SR 3.7.8.2        -------------------------------NOTE-----------------------------
Only required to be performed from June 1 through September 30.
Verify average water temperature is  100 &deg;F.                          In accordance with the Surveillance Frequency Control Program24 hours ANO-1                                                3.7.8-1                          Amendment No. 215,229,
 
ECP 3.7.8 SURVEILLANCE                                                  FREQUENCY Moved from Page 3.7.8-1 SR 3.7.8.2  -------------------------------NOTE-----------------------------
Only required to be performed from June 1 through September 30.
Verify average water temperature is  100 &deg;F.                          In accordance with the Surveillance Frequency Control Program24 hours SR 3.7.8.3  Perform soundings of the ECP to verify:                                In accordance with the Surveillance
: 1. A contained water volume of ECP                                  Frequency Control 70 acre-feet, and                                              Program12 months
: 2. The minimum indicated water level needed to ensure a volume of 70 acre-feet is maintained.
SR 3.7.8.4  Perform visual inspection of the ECP to verify                          In accordance with conformance with design requirements.                                  the Surveillance Frequency Control Program12 months ANO-1                                          3.7.8-2                          Amendment No. 215,229,
 
CREVS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.7.9.1  Operate each CREVS train for  15 minutes.          In accordance with the Surveillance Frequency Control Program31 days SR 3.7.9.2  Perform required CREVS filter testing in            In accordance with accordance with the Ventilation Filter Testing      the VFTP Program (VFTP).
SR 3.7.9.3  Verify the CREVS automatically isolates the Control  In accordance with Room and switches into a recirculation mode of      the Surveillance operation on an actual or simulated actuation        Frequency Control signal.                                              Program18 months SR 3.7.9.4  Perform required CRE unfiltered air inleakage        In accordance with testing in accordance with the Control Room          the Control Room Envelope Habitability Program.                      Envelope Habitability Program.
ANO-1                                  3.7.9-3          Amendment No. 215,221,239,253,
 
CREACS 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR 3.7.10.1  Verify each CREACS train starts, operates for at In accordance with least 1 hour, and maintains control room air    the Surveillance temperature  84 &deg;F D. B.                        Frequency Control Program31 days SR 3.7.10.2  Verify system flow rate of 9900 cfm +/- 10%.      In accordance with the Surveillance Frequency Control Program18 months ANO-1                                  3.7.10-2                Amendment No. 215,
 
PRVS 3.7.11 3.7  PLANT SYSTEMS 3.7.11    Penetration Room Ventilation System (PRVS)
LCO 3.7.11            Two PRVS trains shall be OPERABLE.
                                                -------------------------------------------NOTE----------------------------------------------
The penetration room negative pressure boundary may be opened intermittently under administrative controls.
APPLICABILITY:        MODES 1, 2, 3, and 4.
ACTIONS CONDITION                            REQUIRED ACTION                        COMPLETION TIME A.      One PRVS train                    A.1    Restore PRVS train to                      7 days inoperable.                              OPERABLE status.
B.      Two PRVS trains                  B.1    Restore penetration room                  24 hours inoperable due to                        negative pressure inoperable penetration                  boundary to OPERABLE room negative pressure                  status.
boundary.
C.      Required Action and              C.1    Be in MODE 3.                              6 hours associated Completion Time not met.                    AND OR                              C.2    Be in MODE 5.                              36 hours Both PRVS trains inoperable for reasons other than Condition B.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY Moved to Page 3.7.11-2 SR 3.7.11.1        Operate each PRVS train for  15 minutes.                              In accordance with the Surveillance Frequency Control Program31 days SR 3.7.11.2        Perform required PRVS filter testing in accordance                      In accordance with with the Ventilation Filter Testing Program (VFTP).                    the VFTP ANO-1                                                3.7.11-1                                Amendment No. 215,
 
PRVS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY Moved from Page 3.7.11-1 SR 3.7.11.1  Operate each PRVS train for  15 minutes.            In accordance with the Surveillance Frequency Control Program31 days SR 3.7.11.2  Perform required PRVS filter testing in accordance    In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SURVEILLANCE                                    FREQUENCY SR 3.7.11.3  Verify each PRVS train actuates on an actual or      In accordance with simulated actuation signal.                          the Surveillance Frequency Control Program18 months ANO-1                                  3.7.11-2                      Amendment No. 215,
 
Spent Fuel Pool Water Level 3.7.13 3.7  PLANT SYSTEMS 3.7.13  Spent Fuel Pool Water Level.
LCO 3.7.13              The spent fuel pool water level shall be  23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY:          During movement of irradiated fuel assemblies in the spent fuel pool.
ACTIONS CONDITION                          REQUIRED ACTION                      COMPLETION TIME A. Spent fuel pool water level      A.1  ---------------NOTE--------------
not within limit.                      LCO 3.0.3 is not applicable.
Suspend movement of                    Immediately irradiated fuel assemblies in the spent fuel pool.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                            FREQUENCY SR 3.7.13.1        Verify the spent fuel pool water level is  23 ft above        In accordance with the top of irradiated fuel assemblies seated in the            the Surveillance storage racks.                                                  Frequency Control Program7 days ANO-1                                          3.7.13-1                              Amendment No. 215,
 
Spent Fuel Pool Boron Concentration 3.7.14 3.7  PLANT SYSTEMS 3.7.14    Spent Fuel Pool Boron Concentration LCO 3.7.14            The spent fuel pool boron concentration shall be > 2000 ppm.
APPLICABILITY:        When fuel assemblies are stored in the spent fuel pool.
ACTIONS CONDITION                          REQUIRED ACTION                      COMPLETION TIME A. Spent fuel pool boron          -------------------NOTE------------------
concentration not within      LCO 3.0.3 is not applicable.
limit.                        ---------------------------------------------
A.1      Suspend movement of fuel                Immediately assemblies in the spent fuel pool.
AND A.2      Initiate action to restore              Immediately spent fuel pool boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.7.14.1        Verify the spent fuel pool boron concentration is                  In accordance with
                    > 2000 ppm.                                                        the Surveillance Frequency Control Program7 days ANO-1                                            3.7.14-1                          Amendment No. 215,228,
 
AC Sources - Operating 3.8.1 CONDITION                                REQUIRED ACTION                        COMPLETION TIME F. Required Action and                F.1      Be in MODE 3.                            6 hours Associated Completion Time of Condition A, B, C,        AND D, or E not met.
F.2      ---------------NOTE--------------
LCO 3.0.4.a is not applicable when entering Mode 4.
Be in MODE 4.                            12 hours G. Three or more required            G.1      Enter LCO 3.0.3.                        Immediately AC sources inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                    FREQUENCY SR 3.8.1.1      Verify correct breaker alignment and indicated                          In accordance with power availability for each required offsite circuit.                  the Surveillance Frequency Control Program7 days SR 3.8.1.2      -------------------------------NOTE-----------------------------
All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
Verify each DG starts from standby conditions and,                      In accordance with in  15 seconds achieves ready-to-load                                the Surveillance conditions.                                                            Frequency Control Program31 days ANO-1                                              3.8.1-4                            Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 SURVEILLANCE                                                    FREQUENCY SR 3.8.1.3  ------------------------------NOTES----------------------------
: 1. DG loadings may include gradual loading as recommended by the manufacturer.
: 2. Momentary transients outside the load range do not invalidate this test.
: 3. This Surveillance shall be conducted on only one DG at a time.
: 4. This SR shall be preceded by and follow, without shutdown, a successful performance of SR 3.8.1.2.
Verify each DG is synchronized and loaded and                            In accordance with operates for  60 minutes at a load  2475 kW and                        the Surveillance 2750 kW.                                                                Frequency Control Program31 days SR 3.8.1.4  Verify each day tank contains  160 gallons of fuel                      In accordance with oil.                                                                      the Surveillance Frequency Control Program31 days SR 3.8.1.5  Check for and remove accumulated water from                              In accordance with each day tank.                                                            the Surveillance Frequency Control Program31 days SR 3.8.1.6  Verify the fuel oil transfer system operates to                          In accordance with transfer fuel oil from storage tanks to the day tank.                    the Surveillance Frequency Control Program31 days Moved to Page 3.8.1-6 SR 3.8.1.7  ------------------------------NOTE------------------------------
This Surveillance shall not normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.
ANO-1                                          3.8.1-5                          Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 Moved to Page 3.8.1-6 SURVEILLANCE                                        FREQUENCY Verify automatic transfer of AC power sources to        In accordance with the selected offsite circuit and manual transfer to      the Surveillance the alternate required offsite circuit.                  Frequency Control Program18 months ANO-1                            3.8.1-5                    Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 SURVEILLANCE                                                    FREQUENCY SR 3.8.1.7  ------------------------------NOTE------------------------------
This Surveillance shall not normally be performed in Moved from Page 3.8.1-5 MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.
Verify automatic transfer of AC power sources to                          In accordance with the selected offsite circuit and manual transfer to                      the Surveillance the alternate required offsite circuit.                                  Frequency Control Program18 months SR 3.8.1.8  ------------------------------NOTE------------------------------
All DG starts may be preceded by an engine prelube period.
Verify on an actual or simulated loss of offsite                          In accordance with power signal:                                                            the Surveillance Frequency Control
: a. De-energization of emergency buses;                                Program18 months
: b. Load shedding from emergency buses; and
: c. DG auto-starts from standby condition and:
: 1. achieves ready-to-load conditions in 15 seconds,
: 2. energizes permanently connected loads,
: 3. energizes auto-connected shutdown load through automatic load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
Moved to new Page 3.8.1-7 SR 3.8.1.9  ------------------------------NOTE------------------------------
All DG starts may be preceded by an engine prelube period.
Verify on an actual or simulated loss of offsite                          In accordance with power signal in conjunction with an actual or                            the Surveillance simulated ESF actuation signal:                                          Frequency Control Program18 months
: a. De-energization of emergency buses; ANO-1                                          3.8.1-6                          Amendment No. 215,253,
 
AC Sources - Operating 3.8.1
: b. Load shedding from emergency buses; and Moved to new Page 3.8.1-7
: c. DG auto-starts from standby condition and:
: 1. achieves ready-to-load conditions in 15 seconds,
: 2. energizes permanently connected loads,
: 3. energizes auto-connected emergency loads through load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
ANO-1                            3.8.1-6                    Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 SURVEILLANCE                                                    FREQUENCY SR 3.8.1.9  ------------------------------NOTE------------------------------
All DG starts may be preceded by an engine prelube period.
Verify on an actual or simulated loss of offsite                          In accordance with Moved from Page 3.8.1-6 power signal in conjunction with an actual or                            the Surveillance simulated ESF actuation signal:                                          Frequency Control Program18 months
: a. De-energization of emergency buses;
: b. Load shedding from emergency buses; and
: c. DG auto-starts from standby condition and:
: 1. achieves ready-to-load conditions in 15 seconds,
: 2. energizes permanently connected loads,
: 3. energizes auto-connected emergency loads through load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
ANO-1                                          3.8.1-7                          Amendment No. 215,253,
 
AC Sources - Shutdown 3.8.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.8.2.1  ------------------------------NOTES-----------------------------
: 1. SR 3.8.1.3 is not required to be performed.
: 2. The 15 second acceptance criteria of SR 3.8.1.2 is not applicable.
For AC Sources required to be OPERABLE, the                            In accordance with SRs of Specification 3.8.1, "AC Sources -                              the Surveillance Operating," except SR 3.8.1.4, SR 3.8.1.7,                            Frequency Control SR 3.8.1.8, and SR 3.8.1.9, are applicable.                            Program31 days ANO-1                                          3.8.2-3                              Amendment No. 215,
 
Diesel Fuel Oil and Starting Air 3.8.3 CONDITION                            REQUIRED ACTION                COMPLETION TIME E. Required Action and              E.1    Declare associated DG            Immediately associated Completion                    inoperable.
Time not met.
OR One or more DGs with diesel fuel oil or required starting air subsystem not within limits for reasons other than Condition A, B, C, or D.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.8.3.1        Verify each fuel oil storage tank contains                In accordance with 20,000 gallons of fuel.                                  the Surveillance Frequency Control Program31 days SR 3.8.3.2        Verify fuel oil properties of new and stored fuel oil      In accordance with are tested in accordance with, and maintained              the Diesel Fuel Oil within the limits of, the Diesel Fuel Oil Testing          Testing Program Program.
SR 3.8.3.3        Verify each DG required air start receiver pressure        In accordance with is  175 psig.                                            the Surveillance Frequency Control Program31 days SR 3.8.3.4        Check for and remove accumulated water from                In accordance with each fuel oil storage tank.                                the Surveillance Frequency Control Program31 days ANO-1                                          3.8.3-2                        Amendment No. 215,
 
DC Sources - Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.8.4.1  Verify battery terminal voltage is greater than or                      In accordance with equal to the minimum established float voltage.                        the Surveillance Frequency Control Program7 days SR 3.8.4.2  Verify each battery charger supplies  300 amps at                      In accordance with greater than or equal to the minimum established                        the Surveillance float voltage for  8 hours.                                            Frequency Control Program18 months OR Verify each battery charger can recharge the battery to the fully charged state within 24 hours while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.
SR 3.8.4.3  --------------------------------NOTE-----------------------------
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is adequate to supply, and                      In accordance with maintain in OPERABLE status, the required                              the Surveillance emergency loads for the design duty cycle when                          Frequency Control subjected to a battery service test or a modified                      Program18 months performance discharge test.
ANO-1                                          3.8.4-2                      Amendment No. 215,250,253,
 
Battery Parameters 3.8.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.8.6.1  --------------------------------NOTE-----------------------------
Not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1.
Verify each battery float current is  2 amps.                          In accordance with the Surveillance Frequency Control Program7 days SR 3.8.6.2  Verify each battery pilot cell float voltage is  2.07 V.              In accordance with the Surveillance Frequency Control Program31 days SR 3.8.6.3  Verify each battery connected cell electrolyte level is                In accordance with greater than or equal to minimum established design                    the Surveillance limits.                                                                Frequency Control Program31 days SR 3.8.6.4  Verify each battery pilot cell temperature is greater                  In accordance with than or equal to minimum established design limits.                    the Surveillance Frequency Control Program31 days SR 3.8.6.5  Verify each battery connected cell float voltage is                    In accordance with 2.07 V.                                                              the Surveillance Frequency Control Program92 days Moved to new Page 3.8.6-4 SR 3.8.6.6  --------------------------------NOTE-----------------------------
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is  80% of the                                In accordance with manufacturer's rating when subjected to a                              the Surveillance performance discharge test or a modified                                Frequency Control performance discharge test.                                            Program60 months ANO-1                                          3.8.6-3                          Amendment No. 215,250,
 
Battery Parameters 3.8.6 AND 12 months when battery shows degradation, or has Moved to new Page 3.8.6-4 reached 85% of the expected life with capacity < 100% of manufacturers rating AND 24 months when battery has reached 85% of the expected life with capacity 100% of manufacturers rating ANO-1  3.8.6-3  Amendment No. 215,250,
 
Battery Parameters 3.8.6 SURVEILLANCE                                                  FREQUENCY SR 3.8.6.6  --------------------------------NOTE-----------------------------
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is  80% of the                                In accordance with manufacturer's rating when subjected to a                              the Surveillance performance discharge test or a modified                                Frequency Control performance discharge test.                                            Program60 months Moved from Page 3.8.6-3 AND 12 months when battery shows degradation, or has reached 85% of the expected life with capacity < 100% of manufacturers rating AND 24 months when battery has reached 85% of the expected life with capacity 100% of manufacturers rating ANO-1                                          3.8.6-4                          Amendment No. 215,250,
 
Inverters - Operating 3.8.7 CONDITION                          REQUIRED ACTION                      COMPLETION TIME B. Required Action and            B.1    Be in MODE 3.                        6 hours associated Completion Time not met.                  AND B.2    ---------------NOTE--------------
LCO 3.0.4.a is not applicable when entering Mode 4.
Be in MODE 4.                        12 hours C. Two or more of the four        C.1    Be in MODE 3.                        12 hours inverters required by LCO 3.8.7.a and                AND LCO 3.8.7.b inoperable.
C.2    Be in MODE 5.                        36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.8.7.1      Verify correct inverter voltage, frequency, and                In accordance with alignment to associated 120 VAC buses RS1, RS2,                the Surveillance RS3, and RS4.                                                  Frequency Control Program7 days ANO-1                                        3.8.7-2                    Amendment No. 215,230,253,
 
Inverters - Shutdown 3.8.8 CONDITION                        REQUIRED ACTION              COMPLETION TIME A.  (continued)                  A.2.5    Enter applicable Conditions Immediately and Required Actions of LCO 3.4.11, "Low Temperature Overpressure Protection (LTOP) System,"
for LTOP features made inoperable by AC vital bus inverter(s).
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.8.8.1      Verify correct inverter voltage and alignments to      In accordance with required 120 VAC vital buses.                          the Surveillance Frequency Control Program7 days ANO-1                                        3.8.8-2                    Amendment No. 215,
 
Distribution Systems - Operating 3.8.9 CONDITION                        REQUIRED ACTION                COMPLETION TIME E. Two or more electrical          E.1  Enter LCO 3.0.3.                Immediately power distribution subsystems inoperable that result in a loss of function.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.8.9.1        Verify correct breaker alignments to required AC,        In accordance with DC, and 120 VAC bus electrical power distribution        the Surveillance subsystems.                                              Frequency Control Program7 days ANO-1                                          3.8.9-2            Amendment No. 215,218,230,
 
Distribution Systems - Shutdown 3.8.10 CONDITION                        REQUIRED ACTION                    COMPLETION TIME A.  (continued)                  A.2.2    Suspend movement of              Immediately irradiated fuel assemblies.
AND A.2.3    Suspend operations              Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2.4    Initiate actions to restore      Immediately required AC, DC, and 120 VAC vital bus electrical power distribution subsystems to OPERABLE status.
AND A.2.5    Declare associated              Immediately required decay heat removal subsystem(s) inoperable.
AND A.2.6    Enter applicable Conditions      Immediately and Required Actions of LCO 3.4.11, "Low Temperature Overpressure Protection (LTOP) System,"
for LTOP features made inoperable by Electrical Power Distribution System.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.8.10.1    Verify correct breaker alignments to required AC,          In accordance with DC, and 120 VAC vital bus electrical power                  the Surveillance distribution subsystems.                                    Frequency Control Program7 days ANO-1                                      3.8.10-2                          Amendment No. 215,
 
Boron Concentration 3.9.1 3.9  REFUELING OPERATIONS 3.9.1  Boron Concentration LCO 3.9.1              Boron concentrations of the Reactor Coolant System and the refueling canal shall be maintained within the limit specified in the COLR.
APPLICABILITY:        MODE 6.
                      -------------------------------------------NOTE----------------------------------------------
Only applicable to the refueling canal when connected to the RCS.
ACTIONS CONDITION                              REQUIRED ACTION                          COMPLETION TIME A. Boron concentration not            A.1    Suspend CORE                              Immediately within limit.                              ALTERATIONS.
AND A.2    Suspend positive reactivity                Immediately additions.
AND A.3    Initiate action to restore                Immediately boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.9.1.1          Verify boron concentration is within the limit                          In accordance with specified in the COLR.                                                  the Surveillance Frequency Control Program72 hours ANO-1                                                3.9.1-1                                  Amendment No. 215,
 
Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.9.2.1  Perform CHANNEL CHECK.                                                  In accordance with the Surveillance Frequency Control Program12 hours SR 3.9.2.2  -------------------------------NOTE------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                            In accordance with the Surveillance Frequency Control Program18 months ANO-1                                          3.9.2-2                              Amendment No. 215,
 
Reactor Building Penetrations 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.9.3.1  Verify each required reactor building penetration is                  In accordance with in the required status.                                              the Surveillance Frequency Control Program7 days SR 3.9.3.2  -------------------------------NOTE------------------------------
Not required to be met for reactor building isolation valves and reactor building purge isolation valves in penetrations closed to comply with LCO c.1.
Verify each required reactor building isolation valve                In accordance with and each reactor building purge isolation valve                      the Surveillance actuates to the isolation position.                                  Frequency Control Program18 months SR 3.9.3.3  Perform CHANNEL CALIBRATION of reactor                                In accordance with building purge exhaust radiation monitor.                            the Surveillance Frequency Control Program18 months ANO-1                                          3.9.3-2                            Amendment No. 215,
 
DHR and Coolant Circulation - High Water Level 3.9.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.9.4.1  Verify one DHR loop is in operation.                  In accordance with the Surveillance Frequency Control Program12 hours ANO-1                                  3.9.4-2                      Amendment No. 215,
 
DHR and Coolant Circulation - Low Water Level 3.9.5 CONDITION                        REQUIRED ACTION              COMPLETION TIME B. No DHR loop OPERABLE            B.1    Suspend operations that        Immediately or in operation.                      would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1.
AND B.2    Initiate action to restore one Immediately DHR loop to OPERABLE status and to operation.
AND B.3    Close all reactor building    4 hours penetrations providing direct access from the reactor building atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.9.5.1        Verify one DHR loop is in operation.                    In accordance with the Surveillance Frequency Control Program12 hours SR 3.9.5.2        Verify correct breaker alignment and indicated          In accordance with power available to each required DHR pump.              the Surveillance Frequency Control Program7 days ANO-1                                        3.9.5-2                      Amendment No. 215,
 
Refueling Canal Water Level 3.9.6 3.9  REFUELING OPERATIONS 3.9.6  Refueling Canal Water Level LCO 3.9.6              Refueling canal water level shall be maintained  23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel.
APPLICABILITY:        During movement of irradiated fuel assemblies within the reactor building.
ACTIONS CONDITION                          REQUIRED ACTION                COMPLETION TIME A. Refueling cavity water          A.1    Suspend movement of              Immediately level not within limit.                irradiated fuel assemblies within the reactor building.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.9.6.1          Verify refueling canal water level is  23 feet above      In accordance with the top of irradiated fuel assemblies seated within        the Surveillance the reactor pressure vessel.                              Frequency Control Program24 hours ANO-1                                            3.9.6-1                        Amendment No. 215,
 
Programs and Manuals 5.5 5.0  ADMINSTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.2        Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The program shall include the following:
: a. Preventive maintenance and periodic visual inspection requirements; and
: b. Integrated leak test requirements for each system at least once per 18 months a frequency in accordance with the Surveillance Frequency Control Program. The provisions of SR 3.0.2 are applicable.
5.5.3        Iodine Monitoring This program provides controls that ensure the capability to accurately determine the airborne iodine concentration under accident conditions. The program shall include the following:
: a. Training of personnel;
: b. Procedures for monitoring; and
: c. Provisions for maintenance of sampling and analysis equipment.
5.5.4        Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
: a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
: b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming 10 CFR 20, Appendix B, Table II, Column 2; ANO-1                                        5.0-7                    Amendment No. 215,218,
 
Programs and Manuals 5.5 5.0  ADMINISTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.5        Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
: a. The definition of the CRE and the CRE boundary.
: b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
: c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
: d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASISin accordance with the Surveillance Frequency Control Program. The results shall be trended and used as part of the 18 month assessment of the CRE boundary assessment specified in TS 5.5.5.c.
: e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
: f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
ANO-1                                          5.0-9                Amendment No. 215,239,240,
 
Programs and Manuals 5.5 5.0  ADMINISTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.7        Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. Surface and volumetric examination of the reactor coolant pump flywheels will be conducted coincident with refueling or maintenance shutdowns such that during 10 year intervals all four reactor coolant pump flywheels will be examined. Such examinations will be performed to the extent possible through the access ports, i.e., those areas of the flywheel accessible without motor disassembly. The surface and volumetric examination may be accomplished by Acoustic Emission Examination as an initial examination method. Should the results of the Acoustic Emission Examination indicate that additional examination is necessary to ensure the structural integrity of the flywheel, then other appropriate NDE methods will be performed on the area of concern.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program inspection frequencies.
5.5.8        DELETEDSurveillance Frequency Control Program 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.
ANO-1                                        5.0-11      Amendment No. 215,239,240,250,257,
 
Programs and Manuals 5.5 5.0  ADMINISTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.13    Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:
: a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
: 1. an API gravity or an absolute specific gravity within limits,
: 2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
: 3. water and sediment within limits;
: b. Within 31 days following addition of new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 2D fuel oil;
: c. Total particulate concentration of the fuel oil is  10 mg/l when tested every 31 days based on ASTM D-2276, Method A-2 or A-3 at a Frequency in accordance with the Surveillance Frequency Control Program; and
: d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program surveillance Frequencies.
5.5.14    Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
: a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
: 1. A change in the TS incorporated in the license; or
: 2. A change to the updated SAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
Proposed changes that do meet these criteria shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
: c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the SAR.
ANO-1                                          5.0-17            Amendment No. 215,218,239,250,
 
ATTACHMENT 4 1CAN031801 REVISED TECHNICAL SPECIFICATION PAGES
 
Definitions 1.1 1.1 Definition (continued)
SHUTDOWN MARGIN (SDM)      SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
: a. All full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL RODS must be accounted for in the determination of SDM;
: b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level; and
: c. There is no change in APSR position.
THERMAL POWER              THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
ANO-1                            1.1-5                      Amendment No. 215,218,
 
SDM 3.1.1 3.1  REACTIVITY CONTROL SYSTEMS 3.1.1  SHUTDOWN MARGIN (SDM)
LCO 3.1.1          The SDM shall be within the limit specified in the COLR.
APPLICABILITY:      MODES 3, 4, and 5.
ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME A. SDM not within limit.      A.1  Initiate boration to restore    15 minutes SDM to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.1.1.1        Verify SDM greater than or equal to the limit        In accordance with specified in the COLR.                                the Surveillance Frequency Control Program ANO-1                                      3.1.1-1                        Amendment No. 215,
 
Reactivity Balance 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.1.2.1  -------------------------------NOTES--------------------------
: 1. The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.
: 2.      This Surveillance is not required to be performed prior to entry into MODE 2.
Verify measured core reactivity balance is within                    Once prior to entering
            +/- 1% k/k of predicted values.                                      MODE 1 after each fuel loading AND
                                                                                  ---------NOTE--------
Only required after 60 EFPD In accordance with the Surveillance Frequency Control Program ANO-1                                          3.1.2-2                              Amendment No. 215,
 
CONTROL ROD Group Alignment Limits 3.1.4 CONDITION                      REQUIRED ACTION                      COMPLETION TIME A. (continued)                  A.2.2.3 --------------NOTE-------------
Only required when THERMAL POWER is
                                          > 20% RTP.
Perform SR 3.2.5.1.                  72 hours B. Required Action and          B.1      Be in MODE 3.                      6 hours associated Completion Time for Condition A not met.
C. More than one CONTROL        C.1.1    Verify SDM to be within the        1 hour ROD inoperable, or not                limit provided in the COLR.
aligned within 6.5% of its group average height, or        OR both.
C.1.2    Initiate boration to restore        1 hour SDM to within limit.
AND C.2      Be in MODE 3.                      6 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                          FREQUENCY SR 3.1.4.1      Verify individual CONTROL ROD positions are                In accordance with within 6.5% of their group average height.                  the Surveillance Frequency Control Program SR 3.1.4.2      Verify CONTROL ROD freedom of movement for                  In accordance with each individual CONTROL ROD that is not fully              the Surveillance inserted.                                                  Frequency Control Program ANO-1                                        3.1.4-2                            Amendment No. 215,
 
Safety Rod Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR 3.1.5.1  Verify each safety rod is fully withdrawn. In accordance with the Surveillance Frequency Control Program ANO-1                                    3.1.5-2              Amendment No. 215,
 
APSR Alignment Limits 3.1.6 3.1  REACTIVITY CONTROL SYSTEMS 3.1.6  AXIAL POWER SHAPING ROD (APSR) Alignment Limits LCO 3.1.6            Each APSR shall be OPERABLE and aligned to within 6.5% of its group average height.
APPLICABILITY:        MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION              COMPLETION TIME A. One APSR inoperable, or          A.1    Perform SR 3.2.5.1.      2 hours not aligned to within 6.5%
of its group average height,                                      AND or both.
2 hours after each APSR movement B. Require Action and                B.1    Be in MODE 3              6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.1.6.1          Verify position of each APSR is within 6.5% of the In accordance with group average height.                              the Surveillance Frequency Control Program ANO-1                                          3.1.6-1                    Amendment No. 215,
 
Position Indicator Channels 3.1.7 3.1    REACTIVITY CONTROL SYSTEMS 3.1.7    Position Indicator Channels LCO 3.1.7                  One position indicator channel for each CONTROL ROD and APSR shall be OPERABLE.
APPLICABILITY:              MODES 1 and 2.
ACTIONS
--------------------------------------------------------------NOTES-------------------------------------------------------
Separate Condition entry is allowed for each CONTROL ROD and APSR.
CONDITION                                REQUIRED ACTION                            COMPLETION TIME A. The required position                  A.1        Declare the rod(s)                      Immediately indicator channel                                inoperable.
inoperable for one or more rods.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.1.7.1              Perform CHANNEL CHECK of required position                            In accordance with indicator channel.                                                    the Surveillance Frequency Control Program SR 3.1.7.2              Perform CHANNEL CALIBRATION of required                                In accordance with position indicator channel.                                            the Surveillance Frequency Control Program ANO-1                                                      3.1.7-1                                Amendment No. 215,
 
PHYSICS TESTS Exceptions - MODE 1 3.1.8 CONDITION                      REQUIRED ACTION            COMPLETION TIME B. THERMAL POWER                      B.1  Suspend PHYSICS TESTS    1 hour
    > 85% RTP.                              exceptions.
OR Nuclear overpower trip setpoint > 10% higher than PHYSICS TESTS power level.
OR Nuclear overpower trip setpoint > 90% RTP.
OR
    ------------NOTE-------------
Only required when THERMAL POWER is
    > 20% RTP.
LHR not within limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                            FREQUENCY SR 3.1.8.1          Verify THERMAL POWER is  85% RTP.            In accordance with the Surveillance Frequency Control Program ANO-1                                        3.1.8-2                  Amendment No. 215,
 
PHYSICS TESTS Exceptions - MODE 1 3.1.8 SURVEILLANCE                                                  FREQUENCY SR 3.1.8.2 ------------------------------NOTE----------------------------
Only required when THERMAL POWER is
            > 20% RTP.
Perform SR 3.2.5.1.                                                In accordance with the Surveillance Frequency Control Program SR 3.1.8.3 Verify nuclear overpower trip setpoint  10% RTP                    Within 8 hours prior higher than the THERMAL POWER at which the                          to performance of test is performed, with a maximum setting of                        PHYSICS TESTS at 90% RTP.                                                            each test plateau SR 3.1.8.4 Verify SDM to be within the limits provided in the                  In accordance with COLR.                                                              the Surveillance Frequency Control Program ANO-1                                        3.1.8-3                            Amendment No. 215,
 
PHYSICS TESTS Exceptions - MODE 2 3.1.9 CONDITION                        REQUIRED ACTION              COMPLETION TIME C. Nuclear overpower trip        C.1    Suspend PHYSICS TESTS      1 hour setpoint is not within limit.        exceptions.
OR Nuclear instrumentation high startup rate CONTROL ROD withdrawal inhibit inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.1.9.1        Verify THERMAL POWER is  5% RTP.                In accordance with the Surveillance Frequency Control Program SR 3.1.9.2        Verify nuclear overpower trip setpoint is        Within 8 hours prior 5% RTP.                                        to performance of PHYSICS TESTS SR 3.1.9.3        Verify SDM to be within the limit provided in the In accordance with COLR.                                            the Surveillance Frequency Control Program ANO-1                                      3.1.9-2                    Amendment No. 215,
 
Regulating Rod Insertion Limits 3.2.1 CONDITION                        REQUIRED ACTION                  COMPLETION TIME D. Regulating rod groups          D.1    Initiate boration to restore    15 minutes inserted in unacceptable              SDM to within the limit operation region.                      provided in the COLR.
AND D.2.1 Restore regulating rod            2 hours groups to within restricted operation region.
OR D.2.2 Reduce THERMAL                    2 hours POWER to less than or equal to the THERMAL POWER allowed by the regulating rod group insertion limits.
E. Required Actions and          E.1    Be in MODE 3.                  6 hours associated Completion Times of Conditions C or D not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.2.1.1      Verify regulating rod groups are within the              In accordance with sequence and overlap limits as specified in the          the Surveillance COLR.                                                    Frequency Control Program SR 3.2.1.2      Verify regulating rod groups meet the insertion          In accordance with limits as specified in the COLR.                        the Surveillance Frequency Control Program SR 3.2.1.3      Verify SDM  1% k/k.                                    Within 4 hours prior to achieving criticality ANO-1                                        3.2.1-2                        Amendment No. 215,
 
APSR Insertion Limits 3.2.2 3.2  POWER DISTRIBUTION LIMITS 3.2.2  AXIAL POWER SHAPING ROD (APSR) Insertion Limits LCO 3.2.2            APSRs shall be positioned within the limits specified in the COLR.
APPLICABILITY:      MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                      COMPLETION TIME A. APSRs not within limits.        A.1  -------------NOTE---------------
Only required when THERMAL POWER is
                                        > 20% RTP.
Perform SR 3.2.5.1.                  Once per 2 hours AND A.2  Restore APSRs to within              24 hours limits.
B. Required Action and            B.1  Be in MODE 3.                        6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                          FREQUENCY SR 3.2.2.1        Verify APSRs are within acceptable limits                In accordance with specified in the COLR.                                    the Surveillance Frequency Control Program ANO-1                                        3.2.2-1                            Amendment No. 215,
 
AXIAL POWER IMBALANCE Operating Limits 3.2.3 3.2  POWER DISTRIBUTION LIMITS 3.2.3    AXIAL POWER IMBALANCE Operating Limits LCO 3.2.3            AXIAL POWER IMBALANCE shall be maintained within the limits specified in the COLR.
APPLICABILITY:      MODE 1 with THERMAL POWER > 40% RTP.
ACTIONS CONDITION                      REQUIRED ACTION            COMPLETION TIME A. AXIAL POWER                    A.1  Perform SR 3.2.5.1.      Once per 2 hours IMBALANCE not within limits.                      AND A.2  Reduce AXIAL POWER        24 hours IMBALANCE to within limits.
B. Required Action and            B.1  Reduce THERMAL POWER      4 hours associated Completion              to  40% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.2.3.1        Verify AXIAL POWER IMBALANCE is within limits  In accordance with as specified in the COLR.                      the Surveillance Frequency Control Program ANO-1                                        3.2.3-1                Amendment No. 215,
 
QPT 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.2.4.1  Verify QPT is within limits as specified in the COLR. In accordance with the Surveillance Frequency Control Program AND When QPT has been restored to less than or equal to the steady state limit, 1 hour for 12 consecutive hours, or until verified acceptable at  95% RTP ANO-1                                    3.2.4-3                    Amendment No. 215,
 
RPS Instrumentation 3.3.1 CONDITION                                REQUIRED ACTION                            COMPLETION TIME D. As required by Required                D.1      Be in MODE 3.                            6 hours Action C.1 and referenced in                        AND Table 3.3.1-1.
D.2      Open all control rod drive                6 hours (CRD) trip breakers.
E. As required by Required                E.1      Open all CRD trip                        6 hours Action C.1 and                                  breakers.
referenced in Table 3.3.1-1.
F. As required by Required                F.1      Reduce THERMAL POWER                      6 hours Action C.1 and                                  < 45% RTP.
referenced in Table 3.3.1-1.
G. As required by Required                G.1      Reduce THERMAL POWER                      6 hours Action C.1 and                                  < 10% RTP.
referenced in Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS
-------------------------------------------------------------NOTE--------------------------------------------------------
Refer to Table 3.3.1-1 to determine which SRs apply to each RPS Function.
SURVEILLANCE                                                      FREQUENCY SR 3.3.1.1              Perform CHANNEL CHECK.                                                In accordance with the Surveillance Frequency Control Program ANO-1                                                      3.3.1-2                                Amendment No. 215,
 
RPS Instrumentation 3.3.1 SURVEILLANCE                                                  FREQUENCY SR 3.3.1.2 ------------------------------NOTES--------------------------
: 1. Adjust power range channel output if the absolute difference is > 2% RTP.
: 2. Not required to be performed until 24 hours after THERMAL POWER is  20% RTP.
Compare results of calorimetric heat balance                        In accordance with calculation to power range channel output.                          the Surveillance Frequency Control Program AND Once within 24 hours after a THERMAL POWER change of 10% RTP SR 3.3.1.3 ------------------------------NOTES--------------------------
: 1. Adjust the power range channel imbalance output if the absolute value of the imbalance error is  2% RTP.
: 2. Not required to be performed until 24 hours after THERMAL POWER is  20% RTP.
Compare results of out of core measured AXIAL                      In accordance with POWER IMBALANCE to incore measured AXIAL                            the Surveillance POWER IMBALANCE.                                                    Frequency Control Program SR 3.3.1.4 Perform CHANNEL FUNCTIONAL TEST.                                    In accordance with the Surveillance Frequency Control Program ANO-1                                        3.3.1-3                            Amendment No. 215,
 
RPS Instrumentation 3.3.1 SURVEILLANCE                                                  FREQUENCY SR 3.3.1.5 ------------------------------NOTE----------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                        In accordance with the Surveillance Frequency Control Program ANO-1                                        3.3.1-4                            Amendment No. 215,
 
RPS Instrumentation 3.3.1 Table 3.3.1-1 Reactor Protection System Instrumentation APPLICABLE          CONDITIONS MODES OR          REFERENCED OTHER                FROM SPECIFIED          REQUIRED          SURVEILLANCE                ALLOWABLE FUNCTION                    CONDITIONS          ACTION C.1        REQUIREMENTS                    VALUE
: 1. Nuclear Overpower -
: a. High Setpoint                    1,2(a),3(d)            D                SR 3.3.1.1              104.9% RTP SR 3.3.1.2 SR 3.3.1.4 SR 3.3.1.5
: b. Low Setpoint 2(b),3(b)              E                SR 3.3.1.1                  5% RTP 4(b),5(b)                              SR 3.3.1.4 SR 3.3.1.5
: 2. RCS High Outlet Temperature                1,2                  D                SR 3.3.1.1                  618 &deg;F SR 3.3.1.4 SR 3.3.1.5
: 3. RCS High Pressure                        1,2(a),3(d)            D                SR 3.3.1.1                2355 psig SR 3.3.1.4 SR 3.3.1.5
: 4. RCS Low Pressure                          1,2(a)                D                SR 3.3.1.1                1800 psig SR 3.3.1.4 SR 3.3.1.5
: 5. RCS Variable Low Pressure                  1,2(a)                D                SR 3.3.1.1            As specified in the SR 3.3.1.4                  COLR SR 3.3.1.5
: 6. Reactor Building High Pressure            1,2,3(c)              D                SR 3.3.1.1                18.7 psia SR 3.3.1.4 SR 3.3.1.5
: 7. Reactor Coolant Pump to                    1,2(a)                D                SR 3.3.1.1          55% RTP with one Power                                                                            SR 3.3.1.4        pump operating in each SR 3.3.1.5                    loop.
: 8. Nuclear Overpower RCS Flow                1,2(a)                D                SR 3.3.1.1            As specified in the and Measured AXIAL                                                        SR 3.3.1.3                  COLR POWER IMBALANCE                                                            SR 3.3.1.4 SR 3.3.1.5
: 9. Main Turbine Trip (Oil                  45% RTP                F                SR 3.3.1.1                40.5 psig Pressure)                                                                        SR 3.3.1.4 SR 3.3.1.5
: 10. Loss of Main Feedwater Pumps            10% RTP                G                SR 3.3.1.1                55.5 psig (Control Oil Pressure)                                                    SR 3.3.1.4 SR 3.3.1.5
: 11. Shutdown Bypass RCS High                2(b),3(b)              E                SR 3.3.1.1                1720 psig Pressure                          4(b),5(b)                              SR 3.3.1.4 SR 3.3.1.5 (a)    When not in shutdown bypass operation.
(b)    During shutdown bypass operation with any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(c)    With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(d)    With any CRD trip breaker in the closed position, the CRD system capable of rod withdrawal, and not in shutdown bypass operation.
ANO-1                                                        3.3.1-5                                Amendment No. 215,
 
RPS - RTM 3.3.3 CONDITION                  REQUIRED ACTION            COMPLETION TIME C. Two or more RTMs            C.1  Open all CRD trip breakers. 6 hours inoperable in MODE 4 or 5.
OR OR C.2  Remove power from all      6 hours Required Action and              CRD trip breakers.
associated Completion Time not met in MODE 4 or 5.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR 3.3.3.1      Perform CHANNEL FUNCTIONAL TEST.              In accordance with the Surveillance Frequency Control Program ANO-1                                3.3.3-2                    Amendment No. 215,
 
CRD Trip Devices 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                FREQUENCY SR 3.3.4.1  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program ANO-1                            3.3.4-3      Amendment No. 215,
 
ESAS Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                FREQUENCY SR 3.3.5.1  Perform CHANNEL CHECK.          In accordance with the Surveillance Frequency Control Program SR 3.3.5.2  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3  Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program ANO-1                            3.3.5-2      Amendment No. 215,
 
ESAS Manual Initiation 3.3.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE                  FREQUENCY SR 3.3.6.1  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program ANO-1                            3.3.6-2        Amendment No. 215,
 
ESAS Actuation Logic 3.3.7 3.3    INSTRUMENTATION 3.3.7    Engineered Safeguards Actuation System (ESAS) Actuation Logic LCO 3.3.7                  The ESAS digital actuation logic channels shall be OPERABLE.
APPLICABILITY:              MODES 1 and 2, MODES 3 and 4 when associated engineered safeguards equipment is required to be OPERABLE.
ACTIONS
----------------------------------------------------------NOTE-----------------------------------------------------------
Separate Condition entry is allowed for each digital actuation logic channel.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more digital                        A.1      Place associated                        1 hour actuation logic channels                          component(s) in inoperable.                                        engineered safeguards configuration.
OR A.2      Declare the associated                  1 hour component(s) inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.3.7.1              Perform digital actuation logic CHANNEL                                  In accordance with FUNCTIONAL TEST.                                                        the Surveillance Frequency Control Program ANO-1                                                      3.3.7-1                                Amendment No. 215,
 
DG LOPS 3.3.8 3.3    INSTRUMENTATION 3.3.8    Diesel Generator (DG) Loss of Power Start (LOPS)
LCO 3.3.8                  Two loss of voltage Function relays and two degraded voltage Function relays DG LOPS instrumentation per DG shall be OPERABLE.
APPLICABILITY:              MODES 1, 2, 3, and 4.
ACTIONS
----------------------------------------------------------NOTE----------------------------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A. One or more Functions                      A.1      Restore relay(s) to                      1 hour with one or more relays for                        OPERABLE status.
one or more DGs inoperable.
B. Required Action and                        B.1      Declare affected DG(s)                  Immediately associated Completion                              inoperable.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.3.8.1              Perform CHANNEL CHECK.                                                  In accordance with the Surveillance Frequency Control Program ANO-1                                                      3.3.8-1                                  Amendment No. 215,
 
DG LOPS 3.3.8 SURVEILLANCE                                                  FREQUENCY SR 3.3.8.2 ------------------------------NOTE------------------------------
When DG LOPS instrumentation is placed in an inoperable status solely for performance of this Surveillance, entry into associated Conditions and Required Actions may be delayed up to 4 hours for the loss of voltage Function, provided the one remaining relay monitoring the Function for the bus is OPERABLE.
Perform CHANNEL CALIBRATION with setpoint                            In accordance with Allowable Value as follows:                                          the Surveillance Frequency Control
: a.      Degraded voltage  423.2 V and  436.0 V                      Program with a time delay of 8 seconds +/- 1 second; and
: b.      Loss of voltage  1600 V and  3000 V with a time delay of  0.30 seconds and 0.98 seconds.
ANO-1                                        3.3.8-2                            Amendment No. 215,
 
Source Range Neutron Flux 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.3.9.1  Perform CHANNEL CHECK.                                                In accordance with the Surveillance Frequency Control Program SR 3.3.9.2  ------------------------------NOTE-----------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program ANO-1                                          3.3.9-2                              Amendment No. 215,
 
Intermediate Range Neutron Flux 3.3.10 3.3  INSTRUMENTATION 3.3.10  Intermediate Range Neutron Flux LCO 3.3.10          One intermediate range neutron flux channel shall be OPERABLE.
APPLICABILITY:      MODE 2 MODES 3, 4, and 5 with any control rod drive (CRD) trip breaker in the closed position and the CRD System capable of rod withdrawal.
ACTIONS CONDITION                        REQUIRED ACTION                      COMPLETION TIME A. Required channel              -------------------NOTE-------------------
inoperable.                  Plant temperature changes are allowed provided the temperature change is accounted for in the SDM calculations.
A.1      Suspend operations                  Immediately involving positive reactivity changes.
AND A.2      Open CRD trip breakers.              1 hour SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.3.10.1      Perform CHANNEL CHECK.                                        In accordance with the Surveillance Frequency Control Program SR 3.3.10.2      Perform CHANNEL FUNCTIONAL TEST.                              In accordance with the Surveillance Frequency Control Program ANO-1                                        3.3.10-1                            Amendment No. 215,
 
Intermediate Range Neutron Flux 3.3.10 SURVEILLANCE                                                  FREQUENCY SR 3.3.10.3 ------------------------------NOTE------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program ANO-1                                        3.3.10-2                            Amendment No. 215,
 
EFIC System Instrumentation 3.3.11 CONDITION                                REQUIRED ACTION                          COMPLETION TIME E. Required Action and                        E.1      Reduce THERMAL                            6 hours associated Completion Time                        POWER to  10% RTP.
not met for Function 1.a or 1.d.
F. Required Action and                        F.1      Be in MODE 3.                            6 hours associated Completion Time not met for                        AND Functions 1.c, 2, or 3.
F.2    Reduce steam generator                    12 hours pressure to < 750 psig.
SURVEILLANCE REQUIREMENTS
-----------------------------------------------------------NOTE----------------------------------------------------------
Refer to Table 3.3.11-1 to determine which SRs shall be performed for each EFIC Function.
SURVEILLANCE                                                      FREQUENCY SR 3.3.11.1            Perform CHANNEL CHECK.                                                  In accordance with the Surveillance Frequency Control Program SR 3.3.11.2            Perform CHANNEL FUNCTIONAL TEST.(Notes 1 & 2)                            In accordance with the Surveillance Frequency Control Program SR 3.3.11.3            Perform CHANNEL CALIBRATION.(Notes 1 & 2)                                In accordance with the Surveillance Frequency Control Program ANO-1                                                    3.3.11-2                          Amendment No. 215,227,
 
EFIC System Instrumentation 3.3.11 SURVEILLANCE REQUIREMENTS (continued)
The following notes apply only to the SG Level - Low function:
Note 1: If the as-found channel setpoints are conservative with respect to the Allowable Value but outside their predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoints are not conservative with respect to the Allowable Value, the channel shall be declared inoperable.
Note 2: The instrument channel setpoint(s) shall be reset to a value that is equal to or more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The Limiting Trip Setpoint and the methodology used to determine the Limiting Trip Setpoint and the predefined as-found acceptance criteria band are specified in the Bases.
ANO-1                                          3.3.11-3                  Amendment No. 215,227,
 
EFIC System Instrumentation 3.3.11 Table 3.3.11-1 Emergency Feedwater Initiation and Control System Instrumentation APPLICABLE MODES OR REQUIRED  SURVEILLANCE      ALLOWABLE FUNCTION                  OTHER CHANNELS  REQUIREMENTS          VALUES SPECIFIED CONDITIONS
: 1. EFW Initiation
: a. Loss of MFW Pumps          10% RTP              4      SR 3.3.11.1        55.5 psig (Control Oil Pressure)                                  SR 3.3.11.2 SR 3.3.11.3
: b. SG Level - Low              1,2,3          4 per SG    SR 3.3.11.1    9.34 inches(c,d)
SR 3.3.11.2 SR 3.3.11.3
: c. SG Pressure - Low          1,2,3(a)        4 per SG    SR 3.3.11.1      584.2 psig SR 3.3.11.2 SR 3.3.11.3
: d. RCP Status                10% RTP              4      SR 3.3.11.1          NA SR 3.3.11.2
: 2. EFW Vector Valve Control
: a. SG Pressure - Low          1,2,3(a)        4 per SG    SR 3.3.11.1      584.2 psig SR 3.3.11.2 SR 3.3.11.3
: b. SG Differential            1,2,3(a)            4      SR 3.3.11.1        150 psid Pressure - High                                        SR 3.3.11.2 SR 3.3.11.3
: 3. Main Steam Line Isolation
: a. SG Pressure - Low          1,2,3(a)(b)      4 per SG    SR 3.3.11.1      584.2 psig SR 3.3.11.2 SR 3.3.11.3 (a) When SG pressure  750 psig.
(b) Except when all associated valves are closed and deactivated.
(c) The SG Level - Low Limiting Trip Setpoint in accordance with NRC letter dated September 7, 2005, Technical Specification For Addressing Issues Related To Setpoint Allowable Values, is  10.42 inches.
(d) Includes an actuation time delay of  10.4 seconds.
ANO-1                                        3.3.11-4              Amendment No. 215,227,
 
EFIC Manual Initiation 3.3.12 CONDITION                  REQUIRED ACTION          COMPLETION TIME D. Required Action and          D.1  Be in MODE 3.            6 hours associated Completion Time not met for EFW        AND Initiation Function.
D.2  Be in MODE 4.            12 hours E. Required Action and          E.1  Be in MODE 3.            6 hours associated Completion Time not met for Main        AND Steam Line Isolation Function.                    E.2.1 Reduce steam generator  12 hours pressure to < 750 psig.
OR E.2.2 Close and deactivate all 12 hours associated valves.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                            FREQUENCY SR 3.3.12.1      Perform CHANNEL FUNCTIONAL TEST.            In accordance with the Surveillance Frequency Control Program ANO-1                                  3.3.12-2                  Amendment No. 215,
 
EFIC Logic 3.3.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE                FREQUENCY SR 3.3.13.1  Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program ANO-1                            3.3.13-2      Amendment No. 215,
 
EFIC Vector Logic 3.3.14 3.3  INSTRUMENTATION 3.3.14  Emergency Feedwater Initiation and Control (EFIC) Vector Logic.
LCO 3.3.14        Four channels of the EFIC vector logic shall be OPERABLE.
APPLICABILITY:    MODES 1 and 2, MODE 3 when steam generator pressure is  750 psig.
ACTIONS CONDITION                      REQUIRED ACTION              COMPLETION TIME A. One vector logic channel      A.1    Restore channel to              72 hours inoperable.                        OPERABLE status.
B. Required Action and          B.1    Be in MODE 3.                  6 hours associated Completion Time not met.                AND B.2    Reduce steam generator          12 hours pressure to < 750 psig.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.3.14.1      Perform a CHANNEL FUNCTIONAL TEST.                    In accordance with the Surveillance Frequency Control Program ANO-1                                    3.3.14-1                      Amendment No. 215,
 
PAM Instrumentation 3.3.15 CONDITION                                  REQUIRED ACTION                        COMPLETION TIME E. As required by Required                    E.1      Be in MODE 3.                            6 hours Action D.1 and referenced in Table 3.3.15-1.                      AND E.2      Be in MODE 4.                            12 hours F. As required by Required                    F.1      Initiate action to prepare              Immediately Action D.1 and referenced                          and submit a Special in Table 3.3.15-1.                                Report.
SURVEILLANCE REQUIREMENTS
----------------------------------------------------------NOTE-----------------------------------------------------------
These SRs apply to each PAM instrumentation Function in Table 3.3.15-1.
SURVEILLANCE                                                    FREQUENCY SR 3.3.15.1            Perform CHANNEL CHECK for each required                                  In accordance with instrumentation channel that is normally energized.                      the Surveillance Frequency Control Program SR 3.3.15.2            -------------------------------NOTE------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                            In accordance with the Surveillance Frequency Control Program ANO-1                                                      3.3.15-2                          Amendment No. 215,222,
 
Control Room Isolation - High Radiation 3.3.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.3.16.1  Perform CHANNEL CHECK.                                                In accordance with the Surveillance Frequency Control Program SR 3.3.16.2  -------------------------------NOTE------------------------------
When the Control Room Isolation - High Radiation instrumentation is placed in an inoperable status solely for performance of this Surveillance, entry into associated Conditions and Required Actions may be delayed for up to 3 hours.
Perform CHANNEL FUNCTIONAL TEST.                                      In accordance with the Surveillance Frequency Control Program SR 3.3.16.3  Perform CHANNEL CALIBRATION.                                          In accordance with the Surveillance Frequency Control Program ANO-1                                          3.3.16-2                            Amendment No. 215,
 
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.1    RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1            RCS DNB parameters (loop pressure, hot leg temperature, and RCS total flow rate) shall be within the limits specified in the COLR.
APPLICABILITY:        MODE 1.
                      ------------------------------------------------NOTE-----------------------------------------
RCS loop pressure limit does not apply during pressure transients due to a THERMAL POWER change > 5% RTP per minute.
ACTIONS CONDITION                              REQUIRED ACTION                          COMPLETION TIME A. One or more RCS DNB                A.1      Restore RCS DNB                          2 hours parameters not within                        parameter(s) to within limits.                                      limit.
B. Required Action and                B.1      Be in MODE 2.                            6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.4.1.1        ---------------------------------NOTE----------------------------
With three RCPs operating, the limits are applied to the loop with two RCPs in operation.
Verify RCS loop pressure is within the limit specified                    In accordance with in the COLR.                                                              the Surveillance Frequency Control Program ANO-1                                                3.4.1-1                                Amendment No. 215,
 
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE                                                FREQUENCY SR 3.4.1.2 ---------------------------------NOTE---------------------------
With three RCPs operating, the limits are applied to the loop with two RCPs in operation.
Verify RCS hot leg temperature is within the limit                    In accordance with specified in the COLR.                                                the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow is within the limit specified in                In accordance with the COLR.                                                            the Surveillance Frequency Control Program SR 3.4.1.4 ---------------------------------NOTE---------------------------
Only required to be performed when stable thermal conditions are established at  90% RTP.
Verify RCS total flow rate is within the limit specified              In accordance with in the COLR by measurement.                                          the Surveillance Frequency Control Program ANO-1                                        3.4.1-2                            Amendment No. 215,
 
RCS Minimum Temperature for Criticality 3.4.2 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.2  RCS Minimum Temperature for Criticality LCO 3.4.2              The RCS average temperature (Tavg) shall be  525 &deg;F.
APPLICABILITY:          MODE 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                COMPLETION TIME A. Tavg not within limit.        A.1    Be in MODE 3.                30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.4.2.1          Verify RCS Tavg  525 &deg;F.                            In accordance with the Surveillance Frequency Control Program ANO-1                                        3.4.2-1                      Amendment No. 215,
 
RCS P/T Limits 3.4.3 CONDITION                                REQUIRED ACTION                      COMPLETION TIME D. -------------NOTE-------------          D.1      Initiate action to restore          Immediately Required Action D.2 shall                        parameter(s) to within limit.
be completed whenever this Condition is entered.              AND D.2      Determine RCS is                    Prior to entering Requirements of LCO not                          acceptable for continued            MODE 4 met in other than MODE 1,                        operation.
2, 3, or 4.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.4.3.1          ------------------------------NOTE-----------------------------
Only required to be performed during RCS heatup operations with fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                      In accordance with heatup rates are within the limits specified in                    the Surveillance Figure 3.4.3-1.                                                    Frequency Control Program SR 3.4.3.2          ------------------------------NOTE-----------------------------
Only required to be performed during RCS cooldown operations with fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                      In accordance with cooldown rates are within the limits specified in                  the Surveillance Figure 3.4.3-2.                                                    Frequency Control Program ANO-1                                                  3.4.3-2                            Amendment No. 215,
 
RCS P/T Limits 3.4.3 SURVEILLANCE                                                FREQUENCY SR 3.4.3.3 ------------------------------NOTE-----------------------------
Only required to be performed during RCS heatup and cooldown operations with no fuel in the reactor vessel.
Verify RCS pressure, RCS temperature, and RCS                      In accordance with cooldown rates are within the limits specified in                  the Surveillance Figure 3.4.3-3.                                                    Frequency Control Program SR 3.4.3.4 ------------------------------NOTE----------------------------
Only required to be performed during PHYSICS TESTS with RCS temperature  525 &deg;F.
Verify RCS pressure and RCS temperature are                        In accordance with within the criticality limits specified in Figure 3.4.3-1.          the Surveillance Frequency Control Program ANO-1                                        3.4.3-3                            Amendment No. 215,
 
RCS Loops - MODES 1 and 2 3.4.4 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.4  RCS Loops - MODES 1 and 2 LCO 3.4.4            Two RCS Loops shall be in operation, with:
: a. Four reactor coolant pumps (RCPs) operating; or
: b. Three RCPs operating and THERMAL POWER restricted as specified in the COLR.
APPLICABILITY:      MODES 1 and 2.
ACTIONS CONDITION                        REQUIRED ACTION            COMPLETION TIME A. One RCP not in operation      A.1      Restore one non-operating  18 hours in each loop.                          RCP to operation.
B. Required Action and          B.1      Be in MODE 3.              6 hours associated Completion Time of Condition A not met.
OR LCO not met for reasons other than Condition A.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.4.4.1        Verify required RCS loops are in operation.          In accordance with the Surveillance Frequency Control Program ANO-1                                          3.4.4-1                    Amendment No. 215,
 
RCS Loops - MODE 3 3.4.5 CONDITION                              REQUIRED ACTION                        COMPLETION TIME C. Two RCS loops inoperable. C.1              Suspend operations that                Immediately would cause introduction OR                                        into the RCS, coolant with boron concentration less Required RCS loop not in                  than required to meet SDM operation.                                of LCO 3.1.1.
AND C.2      Initiate action to restore one        Immediately RCS loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.5.1    Verify required RCS loop is in operation.                              In accordance with the Surveillance Frequency Control Program SR 3.4.5.2    --------------------------------NOTE----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required pump.                                the Surveillance Frequency Control Program ANO-1                                            3.4.5-2                              Amendment No. 215,
 
RCS Loops - MODE 4 3.4.6 CONDITION                              REQUIRED ACTION                      COMPLETION TIME B. Two required loops                B.1      Suspend operations that                Immediately inoperable.                                would cause introduction into the RCS, coolant with OR                                          boron concentration less than required to meet SDM Required loop not in                        of LCO 3.1.1.
operation.
AND B.2      Initiate action to restore one          Immediately loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.6.1      Verify required DHR or RCS loop is in operation.                      In accordance with the Surveillance Frequency Control Program SR 3.4.6.2      -------------------------------NOTE-----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required pump.                                the Surveillance Frequency Control Program ANO-1                                            3.4.6-2                                Amendment No. 215,
 
RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.7.1  Verify required DHR loop is in operation.                            In accordance with the Surveillance Frequency Control Program SR 3.4.7.2  Verify required SG secondary side water levels are                    In accordance with 20 inches.                                                          the Surveillance Frequency Control Program SR 3.4.7.3  --------------------------------NOTE-----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required DHR pump.                            the Surveillance Frequency Control Program ANO-1                                          3.4.7-3                            Amendment No. 215,
 
RCS Loops - MODE 5, Loops Not Filled 3.4.8 CONDITION                              REQUIRED ACTION                        COMPLETION TIME B. No required DHR loop              B.1      Suspend operations that              Immediately OPERABLE.                                    would cause introduction into the RCS, coolant with OR                                          boron concentration less than required to meet SDM Required DHR loop not in                    of LCO 3.1.1.
operation.
AND B.2      Suspend all operations                Immediately involving reduction in RCS water volume.
AND B.3      Initiate action to restore one        Immediately DHR loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.8.1      Verify required DHR loop is in operation.                            In accordance with the Surveillance Frequency Control Program SR 3.4.8.2      --------------------------------NOTE-----------------------------
Not required to be performed until 24 hours after a required pump is not in operation.
Verify correct breaker alignment and indicated                        In accordance with power available to each required DHR pump.                            the Surveillance Frequency Control Program ANO-1                                              3.4.8-2                              Amendment No. 215,
 
Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.4.9.1  Verify pressurizer water level  320 inches. In accordance with the Surveillance Frequency Control Program SR 3.4.9.2  Verify capacity of ES bus powered pressurizer  In accordance with heaters  126 kW.                              the Surveillance Frequency Control Program ANO-1                                  3.4.9-2            Amendment No. 215,241,
 
LTOP System 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.4.11.1  Verify pressurizer level does not represent a water                  30 minutes during solid condition.                                                      RCS heatup and cooldown AND In accordance with the Surveillance Frequency Control Program SR 3.4.11.2  Verify HPI is deactivated.                                            In accordance with the Surveillance Frequency Control Program SR 3.4.11.3  Verify each pressurized CFT is isolated.                              In accordance with the Surveillance Frequency Control Program SR 3.4.11.4  --------------------------------NOTE-----------------------------
Verification of locked, sealed, or otherwise secured open vent path(s) only required to be performed every 31 days.
Verify OPERABLE pressure relief capability.                          In accordance with the Surveillance Frequency Control Program SR 3.4.11.5  Perform CHANNEL CALIBRATION of ERV opening                            In accordance with circuitry.                                                            the Surveillance Frequency Control Program ANO-1                                          3.4.11-3                            Amendment No. 215,
 
RCS Specific Activity 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.4.12.1  --------------------------------NOTE-----------------------------
Only required to be performed in MODE 1 and 2, MODE 3 with RCS average temperature  500 &deg;F.
Verify reactor coolant DOSE EQUIVALENT XE-133                        In accordance with specific activity  2200 &#xb5;Ci/gm.                                    the Surveillance Frequency Control Program SR 3.4.12.2  Verify reactor coolant DOSE EQUIVALENT I-131                        In accordance with specific activity  1.0 &#xb5;Ci/gm.                                      the Surveillance Frequency Control Program ANO-1                                          3.4.12-2                    Amendment No. 215,238,243,
 
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.13.1 ------------------------------NOTES-----------------------------
: 1. Not required to be performed until 12 hours after establishment of steady state operation at or near operating pressure.
: 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by                      In accordance with performance of an RCS water inventory balance.                          the Surveillance Frequency Control Program SR 3.4.13.2 ------------------------------NOTE-------------------------------
Not required to be performed until 12 hours after establishment of steady state operation.
Verify primary to secondary LEAKAGE is                                  In accordance with 150 gallons per day through any one SG.                              the Surveillance Frequency Control Program ANO-1                                          3.4.13-2                          Amendment No. 215,224,
 
RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.4.14.1  -------------------------------NOTE------------------------------
Not required to be performed in MODES 3 and 4.
Verify leakage from each RCS pressure isolation                        In accordance with check valve, or pair of check valves, as applicable,                  the INSERVICE is less than or equal to an equivalent of the                          TESTING Allowable Leakage Limit identified below at a                          PROGRAM differential test pressure  150 psid.
AND Pressure Isolation                      Allowable Check Valve(s)                    Leakage Limit                    Once prior to entering MODE 2 DH-14A                              5 gpm                          whenever the unit DH-13A and DH-17                    5 gpm total                    has been in DH-14B                              5 gpm                          MODE 5 for 7 days DH-13B and DH-18                    5 gpm total                    or more, if leakage testing has not been performed in the previous 9 months SR 3.4.14.2  Verify DHR System autoclosure interlock prevents                      In accordance with the valves from being opened with a simulated or                      the Surveillance actual high RCS pressure signal.                                      Frequency Control Program SR 3.4.14.3  Verify DHR System autoclosure interlock causes                        In accordance with the valves to close automatically with a simulated or                  the Surveillance actual high RCS pressure signal:                                      Frequency Control Program
: c. 340 psig for one valve; and
: d. 400 psig for the other valve.
SR 3.4.14.4  Verify DHR System autoclosure interlock prevents                      In accordance with the valves from being opened with a simulated or                      the Surveillance actual Core Flood Tank isolation valve not closed                    Frequency Control signal.                                                                Program SR 3.4.14.5  Verify DHR System autoclosure interlock causes                        In accordance with the valves to close automatically with a simulated or                  the Surveillance actual Core Flood Tank isolation valve not closed                    Frequency Control signal.                                                                Program ANO-1                                          3.4.14-2                          Amendment No. 215,257,
 
RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.4.15.1  Perform CHANNEL CHECK of required reactor          In accordance with building atmosphere radioactivity monitor.        the Surveillance Frequency Control Program SR 3.4.15.2  Perform CHANNEL FUNCTIONAL TEST of                In accordance with required reactor building atmosphere radioactivity the Surveillance monitor.                                          Frequency Control Program SR 3.4.15.3  Perform CHANNEL CALIBRATION of required            In accordance with reactor building atmosphere radioactivity monitor. the Surveillance Frequency Control Program SR 3.4.15.4  Perform CHANNEL CALIBRATION of required            In accordance with reactor building sump monitor.                    the Surveillance Frequency Control Program ANO-1                                  3.4.15-3                  Amendment No. 215,
 
RCS Leakage Detection Instrumentation 3.4.15 3.5  EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1  Core Flood Tanks (CFTs)
LCO 3.5.1              Two CFTs shall be OPERABLE.
APPLICABILITY:        MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) pressure > 800 psig.
ACTIONS CONDITION                          REQUIRED ACTION            COMPLETION TIME A. One CFT inoperable due to A.1        Restore boron                72 hours boron concentration not              concentration to within within limits.                        limits.
B. One CFT inoperable for        B.1    Restore CFT to                1 hour reasons other than                    OPERABLE status.
Condition A.
C. Required Action and            C.1    Be in MODE 3.                6 hours associated Completion Time of Condition A or B      AND not met.
C.2    Reduce RCS pressure to OR                                    800 psig.                  12 hours Two CFTs inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.5.1.1          Verify each CFT isolation valve is fully open.      In accordance with the Surveillance Frequency Control Program ANO-1                                          3.5.1-1                    Amendment No. 215,
 
RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE                                  FREQUENCY SR 3.5.1.2 Verify borated water volume in each CFT is          In accordance with 970 ft3 and  1110 ft3.                          the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each CFT is      In accordance with 560 psig and  640 psig.                          the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each CFT is          In accordance with 2270 ppm.                                        the Surveillance Frequency Control Program AND
                                                                ---------NOTE-------
Only required to be performed for affected CFT Once within 12 hours after each solution level increase of 0.2 feet that is not the result of addition from a borated water source of known concentration 2270 ppm SR 3.5.1.5 Verify power is removed from each CFT isolation    In accordance with valve operator.                                    the Surveillance Frequency Control Program ANO-1                                3.5.1-2                    Amendment No. 215,
 
ECCS - Operating 3.5.2 3.5  EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2  ECCS - Operating LCO 3.5.2            Two ECCS trains shall be OPERABLE.
APPLICABILITY:      MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) temperature > 350 &deg;F.
ACTIONS CONDITION                      REQUIRED ACTION                COMPLETION TIME A. One or more trains            A.1  Restore train(s) to            72 hours inoperable.                        OPERABLE status.
B. Required Action and          B.1  Be in MODE 3.                  6 hours associated Completion Time not met.                AND B.2  Reduce RCS temperature          12 hours to  350 &deg;F.
C. Less than 100% of the        C.1  Enter LCO 3.0.3.                Immediately ECCS flow equivalent to a single OPERABLE train available.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.5.2.1        Verify each ECCS manual, power operated, and            In accordance with automatic valve in the flow path, that is not locked,  the Surveillance sealed, or otherwise secured in position, is in the    Frequency Control correct position.                                      Program ANO-1                                        3.5.2-1                    Amendment No. 215,257,
 
ECCS - Operating 3.5.2 SURVEILLANCE                                    FREQUENCY SR 3.5.2.2 Verify each ECCS pump's developed head at the        In accordance with test flow point is greater than or equal to the      the INSERVICE required developed head.                              TESTING PROGRAM SR 3.5.2.3 Verify each ECCS automatic valve in the flow path    In accordance with that is not locked, sealed, or otherwise secured in  the Surveillance position, actuates to the correct position on an      Frequency Control actual or simulated actuation signal.                Program SR 3.5.2.4 Verify each ECCS pump starts automatically on an      In accordance with actual or simulated actuation signal.                the Surveillance Frequency Control Program SR 3.5.2.5 Verify, by visual inspection, each ECCS train        In accordance with reactor building sump suction inlet is not restricted the Surveillance by debris and screens show no evidence of            Frequency Control structural distress or abnormal corrosion.            Program ANO-1                                  3.5.2-2                    Amendment No. 215,
 
BWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.5.4.1  -------------------------------NOTE-----------------------------
Only required to be performed when ambient air temperature is < 40 &deg;F or > 110 &deg;F.
Verify BWST borated water temperature is                              In accordance with 40 &deg;F and  110 &deg;F.                                                  the Surveillance Frequency Control Program SR 3.5.4.2  Verify BWST borated water level is                                    In accordance with 38.4 feet and  42 feet.                                            the Surveillance Frequency Control Program SR 3.5.4.3  Verify BWST boron concentration is                                    In accordance with 2270 ppm and  2670 ppm.                                            the Surveillance Frequency Control Program ANO-1                                          3.5.4-2                            Amendment No. 215,253,
 
Reactor Building Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.6.2.1  -------------------------------NOTE-----------------------------
: 1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
: 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.
Perform required air lock leakage rate testing in                      In accordance with accordance with the Reactor Building Leakage Rate                      the Reactor Building Testing Program.                                                      Leakage Rate Testing Program SR 3.6.2.2  Verify only one door in the air lock can be opened at                  In accordance with a time.                                                                the Surveillance Frequency Control Program ANO-1                                          3.6.2-4                              Amendment No. 215,
 
Reactor Building Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.6.3.1  Verify each reactor building purge isolation valve is                In accordance with closed.                                                              the Surveillance Frequency Control Program SR 3.6.3.2  -------------------------------NOTE----------------------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each reactor building isolation manual valve                  In accordance with and blind flange that is located outside the reactor                the Surveillance building and not locked, sealed, or otherwise                        Frequency Control secured, and is required to be closed during                        Program accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.
SR 3.6.3.3  -------------------------------NOTE----------------------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each reactor building isolation manual valve                  Prior to entering and blind flange that is located inside the reactor                  MODE 4 from building and not locked, sealed, or otherwise                        MODE 5 if not secured, and required to be closed during accident                  performed within conditions is closed, except for reactor building                    the previous isolation valves that are open under administrative                  92 days controls.
SR 3.6.3.4  Verify the isolation time of each automatic power                    In accordance with operated reactor building isolation valve is within                  the INSERVICE limits.                                                              TESTING PROGRAM SR 3.6.3.5  Verify each automatic reactor building isolation                    In accordance with valve that is not locked, sealed, or otherwise                      the Surveillance secured in position, actuates to the isolation                      Frequency Control position on an actual or simulated actuation signal.                Program ANO-1                                          3.6.3-4                      Amendment No. 215,253,257,
 
Reactor Building Pressure 3.6.4 3.6  REACTOR BUILDING SYSTEMS 3.6.4  Reactor Building Pressure LCO 3.6.4              Reactor building pressure shall be  -1.0 psig and  +3.0 psig.
APPLICABILITY:          MODES 1, 2, 3, and 4.
ACTIONS CONDITION                          REQUIRED ACTION                        COMPLETION TIME A. Reactor building pressure      A.1  Restore reactor building                1 hour not within limits.                    pressure to within limits.
B. Required Action and            B.1  Be in MODE 3.                            6 hours associated Completion Time not met.                  AND B.2  ---------------NOTE--------------
LCO 3.0.4.a is not applicable when entering Mode 4.
Be in MODE 4.                            12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.6.4.1          Verify reactor building pressure is  -1.0 psig and              In accordance with
                      +3.0 psig.                                                    the Surveillance Frequency Control Program ANO-1                                          3.6.4-1                            Amendment No. 215,253,
 
Reactor Building Spray and Cooling System 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.6.5.1  Verify each reactor building spray manual, power          In accordance with operated, and automatic valve in each required flow      the Surveillance path that is not locked, sealed, or otherwise secured    Frequency Control in position is in the correct position.                  Program SR 3.6.5.2  Operate each required reactor building cooling train      In accordance with fan unit for  15 minutes.                                the Surveillance Frequency Control Program SR 3.6.5.3  Verify each required reactor building cooling train      In accordance with cooling water flow rate is  1200 gpm.                    the Surveillance Frequency Control Program SR 3.6.5.4  Verify each required reactor building spray pump's        In accordance with developed head at the flow test point is greater than    the INSERVICE or equal to the required developed head.                  TESTING PROGRAM SR 3.6.5.5  Verify each automatic reactor building spray valve in    In accordance with each required flow path that is not locked, sealed, or    the Surveillance otherwise secured in position, actuates to the            Frequency Control correct position on an actual or simulated actuation      Program signal.
SR 3.6.5.6  Verify each required reactor building spray pump          In accordance with starts automatically on an actual or simulated            the Surveillance actuation signal.                                        Frequency Control Program SR 3.6.5.7  Verify each required reactor building cooling train      In accordance with starts automatically on an actual or simulated            the Surveillance actuation signal.                                        Frequency Control Program SR 3.6.5.8  Verify each spray nozzle is unobstructed.                Following maintenance which could result in nozzle blockage ANO-1                                    3.6.5-3              Amendment No. 215,233,257,
 
Spray Additive System 3.6.6 3.6  REACTOR BUILDING SYSTEMS 3.6.6  Spray Additive System LCO 3.6.6              The Spray Additive System shall be OPERABLE.
APPLICABILITY:        MODES 1 and 2.
ACTIONS CONDITION                          REQUIRED ACTION              COMPLETION TIME A. Spray Additive System          A.1  Restore Spray Additive        72 hours inoperable.                          System to OPERABLE status.
B. Required Action and            B.1  Be in MODE 3.                6 hours associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.6.6.1        Verify each Spray Additive System manual, power        In accordance with operated, and automatic valve in the flow path that    the Surveillance is not locked, sealed, or otherwise secured in        Frequency Control position is in the correct position.                  Program SR 3.6.6.2        Verify sodium hydroxide tank solution volume is        In accordance with 9000 gallons.                                        the Surveillance Frequency Control Program ANO-1                                          3.6.6-1                Amendment No. 215,234,
 
Spray Additive System 3.6.6 SURVEILLANCE                                    FREQUENCY SR 3.6.6.3 Verify sodium hydroxide tank solution concentration    In accordance with is > 6.0 wt% and < 8.5 wt.% NaOH.                      the Surveillance Frequency Control Program SR 3.6.6.4 Verify each Spray Additive System automatic valve      In accordance with in the flow path actuates to the correct position on    the Surveillance an actual or simulated actuation signal.                Frequency Control Program ANO-1                                  3.6.6-2                  Amendment No. 215,234,
 
MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.2.1  --------------------------------NOTE-----------------------------
Only required to be performed in MODES 1 and 2.
Verify isolation time of each MSIV is within the limits                In accordance with specified in the INSERVICE TESTING PROGRAM.                            the INSERVICE TESTING PROGRAM SR 3.7.2.2  --------------------------------NOTE-----------------------------
: 1. Only required to be performed in MODES 1 and 2.
: 2. Not required to be met when SG pressure is
                    < 750 psig.
Verify each MSIV actuates to the isolation position                    In accordance with on an actual or simulated actuation signal.                            the Surveillance Frequency Control Program ANO-1                                          3.7.2-2                          Amendment No. 215,257,
 
MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves 3.7.3 SURVEILLANCE                                                FREQUENCY SR 3.7.3.2 --------------------------------NOTES---------------------------
: 1. Only required to be performed in MODES 1 and 2.
: 2. Not required to be met when SG pressure is
                  < 750 psig.
Verify that each MFIV, Main Feedwater Block Valve,                    In accordance with Low Load Feedwater Control Valve and Startup                          the Surveillance Feedwater Control Valve actuates to the isolation                    Frequency Control position on an actual or simulated actuation signal.                  Program ANO-1                                        3.7.3-3                            Amendment No. 215,
 
Secondary Specific Activity 3.7.4 3.7  PLANT SYSTEMS 3.7.4    Secondary Specific Activity LCO 3.7.4              The specific activity of the secondary coolant shall be  0.1 Ci/gm DOSE EQUIVALENT I-131.
APPLICABILITY:          MODES 1, 2, 3, and 4.
ACTIONS CONDITION                          REQUIRED ACTION                COMPLETION TIME A. Specific activity not within    A.1    Be in MODE 3.                  6 hours limit.
AND A.2    Be in MODE 5.                  36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.7.4.1          Verify the specific activity of the secondary coolant    In accordance with is  0.1 Ci/gm DOSE EQUIVALENT I-131.                    the Surveillance Frequency Control Program ANO-1                                            3.7.4-1                  Amendment No. 215,238,
 
EFW System 3.7.5 CONDITION                              REQUIRED ACTION                      COMPLETION TIME C. Required Action and                C.1      Be in MODE 3.                          6 hours associated Completion Time of Condition A or B          AND not met.
C.2      Be in MODE 4.                          18 hours D. Two EFW trains                    D.1      -------------NOTE----------------
inoperable in MODE 1, 2,                    LCO 3.0.3 and all other or 3.                                      LCO Required Actions requiring MODE changes are suspended until one EFW train is restored to OPERABLE status.
Initiate action to restore              Immediately one EFW train to OPERABLE status.
E. Required EFW train                E.1      Initiate action to restore              Immediately inoperable in MODE 4.                      EFW train to OPERABLE status.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.5.1      Verify each EFW manual, power operated, and                            In accordance with automatic valve in each water flow path and in both                    the Surveillance steam supply flow paths to the steam turbine driven                    Frequency Control pump, that is not locked, sealed, or otherwise                        Program secured in position, is in the correct position.
SR 3.7.5.2      -------------------------------NOTE------------------------------
Not required to be performed for the turbine driven EFW pump, until 24 hours after reaching  750 psig in the steam generators.
Verify the developed head of each EFW pump at                          In accordance with the flow test point is greater than or equal to the                    the INSERVICE required developed head.                                              TESTING PROGRAM ANO-1                                              3.7.5-2                          Amendment No. 215,257,
 
EFW System 3.7.5 SURVEILLANCE                                                FREQUENCY SR 3.7.5.3 -------------------------------NOTE------------------------------
Not required to be met in MODE 4 when steam generator is relied upon for heat removal.
Verify each EFW automatic valve that is not locked,                  In accordance with sealed, or otherwise secured in position, actuates to                the Surveillance the correct position on an actual or simulated                        Frequency Control actuation signal.                                                    Program SR 3.7.5.4 -------------------------------NOTE------------------------------
Not required to be met in MODE 4 when steam generator is relied upon for heat removal.
Verify each EFW pump starts automatically on an                      In accordance with actual or simulated actuation signal.                                the Surveillance Frequency Control Program SR 3.7.5.5 Verify proper alignment of the required EFW flow                      Prior to entering paths by verifying manual valve alignment from the                    MODE 2 whenever Q condensate storage tank to each steam                            the unit has been in generator.                                                            MODE 5, MODE 6, or defueled for a cumulative period of > 30 days SR 3.7.5.6 Verify that feedwater is delivered to each steam                      In accordance with generator using the motor-driven EFW pump.                            the Surveillance Frequency Control Program ANO-1                                        3.7.5-3                            Amendment No. 215,
 
QCST 3.7.6 3.7  PLANT SYSTEMS 3.7.6  Q Condensate Storage Tank (QCST)
LCO 3.7.6            The QCST shall be OPERABLE.
APPLICABILITY:      MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
ACTIONS CONDITION                        REQUIRED ACTION            COMPLETION TIME A. The QCST inoperable.          A.1    Verify by administrative    4 hours means OPERABILITY of backup water supply.        AND Once per 12 hours thereafter AND A.2    Restore QCST to            7 days OPERABLE status.
B. Required Action and          B.1    Be in MODE 3.              6 hours associated Completion Time not met.                AND B.2    Be in MODE 4 without        24 hours reliance on steam generator for heat removal.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.7.6.1        Verify QCST volume is  267,000 gallons when        In accordance with required for both units and  107,000 gallons when  the Surveillance only required for Unit 1.                            Frequency Control Program ANO-1                                        3.7.6-1                    Amendment No. 215,
 
SWS 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.7.7.1  -------------------------------NOTE-----------------------------
Isolation of SWS flow to individual components does not render the SWS inoperable.
Verify each SWS manual, power operated, and                            In accordance with automatic valve in the flow path servicing safety                      the Surveillance related equipment, that is not locked, sealed, or                      Frequency Control otherwise secured in position, is in the correct                      Program position.
SR 3.7.7.2  Verify each SWS automatic valve in the flow path                      In accordance with that is not locked, sealed, or otherwise secured in                    the Surveillance position, actuates to the correct position on an                      Frequency Control actual or simulated actuation signal.                                  Program SR 3.7.7.3  Verify each required SWS pump starts                                  In accordance with automatically on an actual or simulated signal.                        the Surveillance Frequency Control Program ANO-1                                          3.7.7-2                            Amendment No. 215,218,
 
ECP 3.7.8 3.7  PLANT SYSTEMS 3.7.8  Emergency Cooling Pond (ECP)
LCO 3.7.8            The ECP shall be OPERABLE.
APPLICABILITY:        MODES 1, 2, 3, and 4.
ACTIONS CONDITION                        REQUIRED ACTION            COMPLETION TIME A. Degradation of the ECP        A.1    Determine ECP remains        7 days noted pursuant to                    acceptable for continued SR 3.7.8.4 below or by                operation.
other inspection.
B. Required Action and            B.1    Be in MODE 3.                6 hours associated Completion Time of Condition A not        AND met.
B.2    Be in MODE 5.                36 hours OR LCO not met for reasons other than Condition A.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.7.8.1        Verify that the indicated water level of the ECP is    In accordance with greater than or equal to that required for an ECP      the Surveillance volume of 70 acre-ft.                                  Frequency Control Program ANO-1                                          3.7.8-1                  Amendment No. 215,229,
 
ECP 3.7.8 SURVEILLANCE                                                  FREQUENCY SR 3.7.8.2 -------------------------------NOTE-----------------------------
Only required to be performed from June 1 through September 30.
Verify average water temperature is  100 &deg;F.                          In accordance with the Surveillance Frequency Control Program SR 3.7.8.3 Perform soundings of the ECP to verify:                                In accordance with the Surveillance
: 1. A contained water volume of ECP                                  Frequency Control 70 acre-feet, and                                              Program
: 2. The minimum indicated water level needed to ensure a volume of 70 acre-feet is maintained.
SR 3.7.8.4 Perform visual inspection of the ECP to verify                        In accordance with conformance with design requirements.                                  the Surveillance Frequency Control Program ANO-1                                        3.7.8-2                          Amendment No. 215,229,
 
CREVS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.7.9.1  Operate each CREVS train for  15 minutes.          In accordance with the Surveillance Frequency Control Program SR 3.7.9.2  Perform required CREVS filter testing in            In accordance with accordance with the Ventilation Filter Testing      the VFTP Program (VFTP).
SR 3.7.9.3  Verify the CREVS automatically isolates the Control  In accordance with Room and switches into a recirculation mode of      the Surveillance operation on an actual or simulated actuation        Frequency Control signal.                                              Program SR 3.7.9.4  Perform required CRE unfiltered air inleakage        In accordance with testing in accordance with the Control Room          the Control Room Envelope Habitability Program.                      Envelope Habitability Program.
ANO-1                                  3.7.9-3          Amendment No. 215,221,239,253,
 
CREACS 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR 3.7.10.1  Verify each CREACS train starts, operates for at In accordance with least 1 hour, and maintains control room air    the Surveillance temperature  84 &deg;F D. B.                        Frequency Control Program SR 3.7.10.2  Verify system flow rate of 9900 cfm +/- 10%.      In accordance with the Surveillance Frequency Control Program ANO-1                                  3.7.10-2                Amendment No. 215,
 
PRVS 3.7.11 3.7  PLANT SYSTEMS 3.7.11    Penetration Room Ventilation System (PRVS)
LCO 3.7.11            Two PRVS trains shall be OPERABLE.
                      -------------------------------------------NOTE----------------------------------------------
The penetration room negative pressure boundary may be opened intermittently under administrative controls.
APPLICABILITY:        MODES 1, 2, 3, and 4.
ACTIONS CONDITION                              REQUIRED ACTION                          COMPLETION TIME A. One PRVS train                    A.1    Restore PRVS train to                      7 days inoperable.                                OPERABLE status.
B. Two PRVS trains                    B.1    Restore penetration room                  24 hours inoperable due to                          negative pressure inoperable penetration                    boundary to OPERABLE room negative pressure                    status.
boundary.
C. Required Action and                C.1    Be in MODE 3.                              6 hours associated Completion Time not met.                      AND OR                                C.2    Be in MODE 5.                              36 hours Both PRVS trains inoperable for reasons other than Condition B.
ANO-1                                              3.7.11-1                                Amendment No. 215,
 
PRVS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.7.11.1  Operate each PRVS train for  15 minutes.          In accordance with the Surveillance Frequency Control Program SR 3.7.11.2  Perform required PRVS filter testing in accordance  In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.11.3  Verify each PRVS train actuates on an actual or    In accordance with simulated actuation signal.                        the Surveillance Frequency Control Program ANO-1                                    3.7.11-2                Amendment No. 215,
 
Spent Fuel Pool Water Level 3.7.13 3.7  PLANT SYSTEMS 3.7.13  Spent Fuel Pool Water Level.
LCO 3.7.13              The spent fuel pool water level shall be  23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY:          During movement of irradiated fuel assemblies in the spent fuel pool.
ACTIONS CONDITION                          REQUIRED ACTION                      COMPLETION TIME A. Spent fuel pool water level      A.1  ---------------NOTE--------------
not within limit.                      LCO 3.0.3 is not applicable.
Suspend movement of                    Immediately irradiated fuel assemblies in the spent fuel pool.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                            FREQUENCY SR 3.7.13.1        Verify the spent fuel pool water level is  23 ft above        In accordance with the top of irradiated fuel assemblies seated in the            the Surveillance storage racks.                                                  Frequency Control Program ANO-1                                          3.7.13-1                              Amendment No. 215,
 
Spent Fuel Pool Boron Concentration 3.7.14 3.7  PLANT SYSTEMS 3.7.14    Spent Fuel Pool Boron Concentration LCO 3.7.14            The spent fuel pool boron concentration shall be > 2000 ppm.
APPLICABILITY:        When fuel assemblies are stored in the spent fuel pool.
ACTIONS CONDITION                          REQUIRED ACTION                      COMPLETION TIME B. Spent fuel pool boron          -------------------NOTE------------------
concentration not within      LCO 3.0.3 is not applicable.
limit.                        ---------------------------------------------
A.1      Suspend movement of fuel                Immediately assemblies in the spent fuel pool.
AND A.2      Initiate action to restore              Immediately spent fuel pool boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.7.14.1        Verify the spent fuel pool boron concentration is                  In accordance with
                    > 2000 ppm.                                                        the Surveillance Frequency Control Program ANO-1                                            3.7.14-1                          Amendment No. 215,228,
 
AC Sources - Operating 3.8.1 CONDITION                                REQUIRED ACTION                        COMPLETION TIME F. Required Action and                F.1      Be in MODE 3.                            6 hours Associated Completion Time of Condition A, B, C,        AND D, or E not met.
F.2      ---------------NOTE--------------
LCO 3.0.4.a is not applicable when entering Mode 4.
Be in MODE 4.                            12 hours G. Three or more required            G.1      Enter LCO 3.0.3.                        Immediately AC sources inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                    FREQUENCY SR 3.8.1.1      Verify correct breaker alignment and indicated                          In accordance with power availability for each required offsite circuit.                  the Surveillance Frequency Control Program SR 3.8.1.2      -------------------------------NOTE-----------------------------
All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
Verify each DG starts from standby conditions and,                      In accordance with in  15 seconds achieves ready-to-load                                the Surveillance conditions.                                                            Frequency Control Program ANO-1                                              3.8.1-4                            Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 SURVEILLANCE                                                    FREQUENCY SR 3.8.1.3 ------------------------------NOTES----------------------------
: 1. DG loadings may include gradual loading as recommended by the manufacturer.
: 2. Momentary transients outside the load range do not invalidate this test.
: 3. This Surveillance shall be conducted on only one DG at a time.
: 4. This SR shall be preceded by and follow, without shutdown, a successful performance of SR 3.8.1.2.
Verify each DG is synchronized and loaded and                          In accordance with operates for  60 minutes at a load  2475 kW and                      the Surveillance 2750 kW.                                                              Frequency Control Program SR 3.8.1.4 Verify each day tank contains  160 gallons of fuel                    In accordance with oil.                                                                    the Surveillance Frequency Control Program SR 3.8.1.5 Check for and remove accumulated water from                            In accordance with each day tank.                                                          the Surveillance Frequency Control Program SR 3.8.1.6 Verify the fuel oil transfer system operates to                        In accordance with transfer fuel oil from storage tanks to the day tank.                  the Surveillance Frequency Control Program ANO-1                                        3.8.1-5                            Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 SURVEILLANCE                                                    FREQUENCY SR 3.8.1.7 ------------------------------NOTE------------------------------
This Surveillance shall not normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.
Verify automatic transfer of AC power sources to                        In accordance with the selected offsite circuit and manual transfer to                    the Surveillance the alternate required offsite circuit.                                Frequency Control Program SR 3.8.1.8 ------------------------------NOTE------------------------------
All DG starts may be preceded by an engine prelube period.
Verify on an actual or simulated loss of offsite                        In accordance with power signal:                                                          the Surveillance Frequency Control
: a. De-energization of emergency buses;                              Program
: b. Load shedding from emergency buses; and
: c. DG auto-starts from standby condition and:
: 1. achieves ready-to-load conditions in 15 seconds,
: 2. energizes permanently connected loads,
: 3. energizes auto-connected shutdown load through automatic load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
ANO-1                                        3.8.1-6                            Amendment No. 215,253,
 
AC Sources - Operating 3.8.1 SURVEILLANCE                                                    FREQUENCY SR 3.8.1.9 ------------------------------NOTE------------------------------
All DG starts may be preceded by an engine prelube period.
Verify on an actual or simulated loss of offsite                        In accordance with power signal in conjunction with an actual or                          the Surveillance simulated ESF actuation signal:                                        Frequency Control Program
: a. De-energization of emergency buses;
: b. Load shedding from emergency buses; and
: c. DG auto-starts from standby condition and:
: 1. achieves ready-to-load conditions in 15 seconds,
: 2. energizes permanently connected loads,
: 3. energizes auto-connected emergency loads through load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
ANO-1                                        3.8.1-7                            Amendment No. 215,253,
 
AC Sources - Shutdown 3.8.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.8.2.1  ------------------------------NOTES-----------------------------
: 1. SR 3.8.1.3 is not required to be performed.
: 2. The 15 second acceptance criteria of SR 3.8.1.2 is not applicable.
For AC Sources required to be OPERABLE, the                            In accordance with SRs of Specification 3.8.1, "AC Sources -                              the Surveillance Operating," except SR 3.8.1.4, SR 3.8.1.7,                            Frequency Control SR 3.8.1.8, and SR 3.8.1.9, are applicable.                            Program ANO-1                                          3.8.2-3                              Amendment No. 215,
 
Diesel Fuel Oil and Starting Air 3.8.3 CONDITION                            REQUIRED ACTION                COMPLETION TIME E. Required Action and              E.1    Declare associated DG            Immediately associated Completion                    inoperable.
Time not met.
OR One or more DGs with diesel fuel oil or required starting air subsystem not within limits for reasons other than Condition A, B, C, or D.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.8.3.1        Verify each fuel oil storage tank contains                In accordance with 20,000 gallons of fuel.                                  the Surveillance Frequency Control Program SR 3.8.3.2        Verify fuel oil properties of new and stored fuel oil      In accordance with are tested in accordance with, and maintained              the Diesel Fuel Oil within the limits of, the Diesel Fuel Oil Testing          Testing Program Program.
SR 3.8.3.3        Verify each DG required air start receiver pressure        In accordance with is  175 psig.                                            the Surveillance Frequency Control Program SR 3.8.3.4        Check for and remove accumulated water from                In accordance with each fuel oil storage tank.                                the Surveillance Frequency Control Program ANO-1                                          3.8.3-2                        Amendment No. 215,
 
DC Sources - Operating SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.8.4.1  Verify battery terminal voltage is greater than or                      In accordance with equal to the minimum established float voltage.                        the Surveillance Frequency Control Program SR 3.8.4.2  Verify each battery charger supplies  300 amps at                      In accordance with greater than or equal to the minimum established                        the Surveillance float voltage for  8 hours.                                            Frequency Control Program OR Verify each battery charger can recharge the battery to the fully charged state within 24 hours while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.
SR 3.8.4.3  --------------------------------NOTE-----------------------------
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is adequate to supply, and                      In accordance with maintain in OPERABLE status, the required                              the Surveillance emergency loads for the design duty cycle when                          Frequency Control subjected to a battery service test or a modified                      Program performance discharge test.
ANO-1                                          3.8.4-2                      Amendment No. 215,250,253,
 
Battery Parameters 3.8.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.8.6.1  --------------------------------NOTE-----------------------------
Not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1.
Verify each battery float current is  2 amps.                          In accordance with the Surveillance Frequency Control Program SR 3.8.6.2  Verify each battery pilot cell float voltage is  2.07 V.              In accordance with the Surveillance Frequency Control Program SR 3.8.6.3  Verify each battery connected cell electrolyte level is                In accordance with greater than or equal to minimum established design                    the Surveillance limits.                                                                Frequency Control Program SR 3.8.6.4  Verify each battery pilot cell temperature is greater                  In accordance with than or equal to minimum established design limits.                    the Surveillance Frequency Control Program SR 3.8.6.5  Verify each battery connected cell float voltage is                    In accordance with 2.07 V.                                                              the Surveillance Frequency Control Program ANO-1                                          3.8.6-3                          Amendment No. 215,250,
 
Battery Parameters 3.8.6 SURVEILLANCE                                                  FREQUENCY SR 3.8.6.6 --------------------------------NOTE-----------------------------
This Surveillance shall not be performed in MODE 1, 2, 3, or 4. However, credit may be taken for unplanned events that satisfy this SR.
Verify battery capacity is  80% of the                                In accordance with manufacturer's rating when subjected to a                              the Surveillance performance discharge test or a modified                                Frequency Control performance discharge test.                                            Program AND 12 months when battery shows degradation, or has reached 85% of the expected life with capacity
                                                                                    < 100% of manufacturers rating AND 24 months when battery has reached 85% of the expected life with capacity  100% of manufacturers rating ANO-1                                          3.8.6-4                          Amendment No. 215,250,
 
Inverters - Operating 3.8.7 CONDITION                          REQUIRED ACTION                      COMPLETION TIME B. Required Action and            B.1    Be in MODE 3.                        6 hours associated Completion Time not met.                  AND B.2    ---------------NOTE--------------
LCO 3.0.4.a is not applicable when entering Mode 4.
Be in MODE 4.                        12 hours C. Two or more of the four        C.1    Be in MODE 3.                        12 hours inverters required by LCO 3.8.7.a and                AND LCO 3.8.7.b inoperable.
C.2    Be in MODE 5.                        36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY SR 3.8.7.1      Verify correct inverter voltage, frequency, and                In accordance with alignment to associated 120 VAC buses RS1, RS2,                the Surveillance RS3, and RS4.                                                  Frequency Control Program ANO-1                                        3.8.7-2                    Amendment No. 215,230,253,
 
Inverters - Shutdown 3.8.8 CONDITION                        REQUIRED ACTION              COMPLETION TIME A.  (continued)                  A.2.5    Enter applicable Conditions Immediately and Required Actions of LCO 3.4.11, "Low Temperature Overpressure Protection (LTOP) System,"
for LTOP features made inoperable by AC vital bus inverter(s).
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.8.8.1      Verify correct inverter voltage and alignments to      In accordance with required 120 VAC vital buses.                          the Surveillance Frequency Control Program ANO-1                                        3.8.8-2                    Amendment No. 215,
 
Distribution Systems - Operating 3.8.9 CONDITION                        REQUIRED ACTION                COMPLETION TIME E. Two or more electrical          E.1  Enter LCO 3.0.3.                Immediately power distribution subsystems inoperable that result in a loss of function.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.8.9.1        Verify correct breaker alignments to required AC,        In accordance with DC, and 120 VAC bus electrical power distribution        the Surveillance subsystems.                                              Frequency Control Program ANO-1                                          3.8.9-2            Amendment No. 215,218,230,
 
Distribution Systems - Shutdown 3.8.10 CONDITION                        REQUIRED ACTION                    COMPLETION TIME A.  (continued)                  A.2.2    Suspend movement of              Immediately irradiated fuel assemblies.
AND A.2.3    Suspend operations              Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2.4    Initiate actions to restore      Immediately required AC, DC, and 120 VAC vital bus electrical power distribution subsystems to OPERABLE status.
AND A.2.5    Declare associated              Immediately required decay heat removal subsystem(s) inoperable.
AND A.2.6    Enter applicable Conditions      Immediately and Required Actions of LCO 3.4.11, "Low Temperature Overpressure Protection (LTOP) System,"
for LTOP features made inoperable by Electrical Power Distribution System.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.8.10.1    Verify correct breaker alignments to required AC,          In accordance with DC, and 120 VAC vital bus electrical power                  the Surveillance distribution subsystems.                                    Frequency Control Program ANO-1                                      3.8.10-2                          Amendment No. 215,
 
Boron Concentration 3.9.1 3.9  REFUELING OPERATIONS 3.9.1  Boron Concentration LCO 3.9.1              Boron concentrations of the Reactor Coolant System and the refueling canal shall be maintained within the limit specified in the COLR.
APPLICABILITY:        MODE 6.
                      -------------------------------------------NOTE----------------------------------------------
Only applicable to the refueling canal when connected to the RCS.
ACTIONS CONDITION                              REQUIRED ACTION                          COMPLETION TIME A. Boron concentration not            A.1    Suspend CORE                              Immediately within limit.                              ALTERATIONS.
AND A.2    Suspend positive reactivity                Immediately additions.
AND A.3    Initiate action to restore                Immediately boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.9.1.1          Verify boron concentration is within the limit                          In accordance with specified in the COLR.                                                  the Surveillance Frequency Control Program ANO-1                                                3.9.1-1                                  Amendment No. 215,
 
Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY SR 3.9.2.1  Perform CHANNEL CHECK.                                                  In accordance with the Surveillance Frequency Control Program SR 3.9.2.2  -------------------------------NOTE------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION.                                            In accordance with the Surveillance Frequency Control Program ANO-1                                          3.9.2-2                              Amendment No. 215,
 
Reactor Building Penetrations 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                FREQUENCY SR 3.9.3.1  Verify each required reactor building penetration is                  In accordance with in the required status.                                              the Surveillance Frequency Control Program SR 3.9.3.2  -------------------------------NOTE------------------------------
Not required to be met for reactor building isolation valves and reactor building purge isolation valves in penetrations closed to comply with LCO c.1.
Verify each required reactor building isolation valve                In accordance with and each reactor building purge isolation valve                      the Surveillance actuates to the isolation position.                                  Frequency Control Program SR 3.9.3.3  Perform CHANNEL CALIBRATION of reactor                                In accordance with building purge exhaust radiation monitor.                            the Surveillance Frequency Control Program ANO-1                                          3.9.3-2                            Amendment No. 215,
 
DHR and Coolant Circulation - High Water Level 3.9.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.9.4.1  Verify one DHR loop is in operation.                  In accordance with the Surveillance Frequency Control Program ANO-1                                  3.9.4-2                      Amendment No. 215,
 
DHR and Coolant Circulation - Low Water Level 3.9.5 CONDITION                        REQUIRED ACTION              COMPLETION TIME B. No DHR loop OPERABLE            B.1    Suspend operations that        Immediately or in operation.                      would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1.
AND B.2    Initiate action to restore one Immediately DHR loop to OPERABLE status and to operation.
AND B.3    Close all reactor building    4 hours penetrations providing direct access from the reactor building atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.9.5.1        Verify one DHR loop is in operation.                    In accordance with the Surveillance Frequency Control Program SR 3.9.5.2        Verify correct breaker alignment and indicated          In accordance with power available to each required DHR pump.              the Surveillance Frequency Control Program ANO-1                                        3.9.5-2                      Amendment No. 215,
 
Refueling Canal Water Level 3.9.6 3.9  REFUELING OPERATIONS 3.9.6  Refueling Canal Water Level LCO 3.9.6              Refueling canal water level shall be maintained  23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel.
APPLICABILITY:        During movement of irradiated fuel assemblies within the reactor building.
ACTIONS CONDITION                          REQUIRED ACTION                COMPLETION TIME A. Refueling cavity water          A.1    Suspend movement of              Immediately level not within limit.                irradiated fuel assemblies within the reactor building.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.9.6.1          Verify refueling canal water level is  23 feet above      In accordance with the top of irradiated fuel assemblies seated within        the Surveillance the reactor pressure vessel.                              Frequency Control Program ANO-1                                            3.9.6-1                        Amendment No. 215,
 
Programs and Manuals 5.5 5.0  ADMINSTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.2        Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The program shall include the following:
: a. Preventive maintenance and periodic visual inspection requirements; and
: b. Integrated leak test requirements for each system at a frequency in accordance with the Surveillance Frequency Control Program. The provisions of SR 3.0.2 are applicable.
5.5.3        Iodine Monitoring This program provides controls that ensure the capability to accurately determine the airborne iodine concentration under accident conditions. The program shall include the following:
: a. Training of personnel;
: b. Procedures for monitoring; and
: c. Provisions for maintenance of sampling and analysis equipment.
5.5.4        Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
: a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
: b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming 10 CFR 20, Appendix B, Table II, Column 2; ANO-1                                        5.0-7                    Amendment No. 215,218,
 
Programs and Manuals 5.5 5.0  ADMINISTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.5        Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
: a. The definition of the CRE and the CRE boundary.
: b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
: c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
: d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency in accordance with the Surveillance Frequency Control Program. The results shall be trended and used as part of the CRE boundary assessment specified in TS 5.5.5.c.
: e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
: f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
ANO-1                                          5.0-9                Amendment No. 215,239,240,
 
Programs and Manuals 5.5 5.0  ADMINISTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.7        Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. Surface and volumetric examination of the reactor coolant pump flywheels will be conducted coincident with refueling or maintenance shutdowns such that during 10 year intervals all four reactor coolant pump flywheels will be examined. Such examinations will be performed to the extent possible through the access ports, i.e., those areas of the flywheel accessible without motor disassembly. The surface and volumetric examination may be accomplished by Acoustic Emission Examination as an initial examination method. Should the results of the Acoustic Emission Examination indicate that additional examination is necessary to ensure the structural integrity of the flywheel, then other appropriate NDE methods will be performed on the area of concern.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program inspection frequencies.
5.5.8        Surveillance Frequency Control Program 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.
ANO-1                                        5.0-11      Amendment No. 215,239,240,250,257,
 
Programs and Manuals 5.5 5.0  ADMINISTRATIVE CONTROLS 5.5  Programs and Manuals 5.5.13    Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:
: a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
: 1. an API gravity or an absolute specific gravity within limits,
: 2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
: 3. water and sediment within limits;
: b. Within 31 days following addition of new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 2D fuel oil;
: c. Total particulate concentration of the fuel oil is  10 mg/l when tested based on ASTM D-2276, Method A-2 or A-3 at a Frequency in accordance with the Surveillance Frequency Control Program; and
: d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program surveillance Frequencies.
5.5.14    Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
: a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
: 1. A change in the TS incorporated in the license; or
: 2. A change to the updated SAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
Proposed changes that do meet these criteria shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
: c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the SAR.
ANO-1                                          5.0-17            Amendment No. 215,218,239,250,
 
ATTACHMENT 5 1CAN031801 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (Information only)
 
SDM B 3.1.1 SURVEILLANCE REQUIREMENTS (continued)
SR 3.1.1.1 (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.The Frequency of 24 hours is based on the generally slow change in required boron concentration, and also allows sufficient time for the operator to collect the required data, which may include performing a boron concentration analysis, and complete the calculation.
REFERENCES
: 1. SAR, Section 1.4, GDC 26.
: 2. SAR, Chapter 3.
: 3. 10 CFR 50.36.
ANO-1                                        B 3.1.1-4                          Amendment No. 215 Rev.
 
Reactivity Balance B 3.1.2 ACTIONS (continued)
A.1 and A.2 (continued)
If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the appropriate reactivity parameter may be renormalized, and operation in MODE 1 may continue. If operational restrictions or additional surveillance requirements are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.
The required Completion Time of 7 days is adequate for preparing operating restrictions or surveillances that may be required to allow continued reactor operation.
B.1 If the core reactivity balance cannot be restored to within the +/-1% k/k limit, the unit must be brought to a MODE in which the LCO does not apply. As a conservative measure, the unit must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then boration required by Required Action A.1 of LCO 3.1.1 would occur. The allowed Completion Time of 6 hours is reasonable, based on operating experience to reach the required unit conditions from RTP in an orderly manner and without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.1.2.1 Core reactivity is verified by a periodic reactivity balance calculation that compares the predicted core reactivity to the actual core reactivity condition (net reactivity of zero condition). The comparison is made considering that core conditions are fixed or stable, including CONTROL ROD and APSR positions, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed once prior to entering MODE 1 after each fuel loading as an initial check on core reactivity conditions and design calculations at BOC. A Note is included in the SR to indicate that the normalization of predicted core reactivity to the measured value may take place within the first 60 effective full power days (EFPD) after each fuel loading. The required Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1 is acceptable, based on the slow rate of core reactivity changes due to fuel depletion and the presence of other indicators (QPT, etc.) for prompt indication of an anomaly. The 60 EFPD after entering MODE 1 allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Another Note is included in the SRs to indicate that the performance of the Surveillance is not required for entry into MODE 2.
ANO-1                                          B 3.1.2-4                          Amendment No. 215 Rev.
 
CONTROL ROD Group Alignment Limits B 3.1.4 SURVEILLANCE REQUIREMENTS SR 3.1.4.1 Verification that individual CONTROL RODS are aligned within 6.5% of their group average height limits at a 12 hour Frequency allows the operator to detect a rod that is beginning to deviate from its expected position. The Surveillancespecified Frequency is controlled under the Surveillance Frequency Control Programtakes into account other CONTROL ROD position information that is continuously available to the operator in the control room, so that during actual CONTROL ROD motion, deviations can immediately be detected.
SR 3.1.4.2 Verifying each CONTROL ROD is OPERABLE would require that each rod be tripped.
However, in MODES 1 and 2, tripping each CONTROL ROD could result in radial tilts.
Exercising each individual CONTROL ROD every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each CONTROL ROD by approximately 1.5% (approximately 2 inches) will not cause radial or axial power tilts, or oscillations, to occur. No additional allowances for instrument uncertainty are required to be incorporated in the implementing procedures for this parameter. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. Between typical performances of SR 3.1.4.2 (determination of CONTROL ROD OPERABILITY by movement), if a CONTROL ROD(S) is discovered to be immovable, but is otherwise determined to be capable of being fully inserted, the CONTROL ROD(S) may continue to be considered OPERABLE unless inoperable for some other reason. At any time, if a CONTROL ROD(S) is immovable, a determination of the capability to fully insert (OPERABILITY) the CONTROL ROD(S) must be made and appropriate action taken. Provided the CONTROL ROD(S) are capable of full insertion, no further assessment is necessary.
When one or more CONTROL ROD(S) have failed to fully insert, additional assessment of CONTROL ROD OPERABILITY is required. Any CONTROL ROD that fails to fully insert may be considered OPERABLE provided:
: a. The CONTROL ROD is  90% inserted or, during power operation, a determination has been made that the CONTROL ROD will insert to at least 90% upon a reactor trip.
: b. SDM is verified. The SDM shall assume the affected CONTROL ROD remains 10%
withdrawn. This SDM correction for the affected CONTROL ROD(S) shall be included in appropriate reactivity control procedures for the remaining reactor operational cycle or until the condition causing the failure to fully insert is eliminated.
: c. The cause or probable cause of the condition, if left uncorrected, provides reasonable assurance that the CONTROL ROD will insert to  90% inserted upon future reactor trip.
ANO-1                                          B 3.1.4-7                          Amendment No. 215 Rev. 13,25,
 
Safety Rod Insertion Limit B 3.1.5 SURVEILLANCE REQUIREMENTS SR 3.1.5.1 Verification that each safety rod is fully withdrawn ensures the safety rods are available to provide reactor shutdown capability.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramVerification that individual safety rod positions are fully withdrawn at a 12 hour Frequency allows the operator to detect a safety rod beginning to deviate from its expected position. Also, the 12 hour Frequency takes into account other information available in the control room for the purpose of monitoring the status of the safety rods.
REFERENCES
: 1. SAR, Section 1.4, GDC 10, GDC 26, and GDC 28.
: 2. 10 CFR 50.46.
: 3. SAR, Chapters 3 and 4.
: 4. BAW-10179P-A, Safety Criteria and Methodology for Acceptable Cycle Reload Analyses, Rev. 2.
: 5. 10 CFR 50.36.
ANO-1                                          B 3.1.5-4                      Amendment No. 215 Rev.
 
APSR Alignment Limits B 3.1.6 ACTIONS (continued)
A.1 (continued)
An alternate to realigning a single misaligned APSR to the group average position is to align the remainder of the APSR group to the position of the misaligned or inoperable APSR, while maintaining APSR insertion, in accordance with the limits in the COLR. This restores the alignment requirements. Deviations up to 2 hours will not cause significant xenon redistribution to occur. This alternative assumes the APSR group movement does not cause the limits of LCO 3.2.2, AXIAL POWER SHAPING ROD (APSR) Insertion Limits, to be exceeded. For this reason, APSR group movement is only practical for instances where small movements of the APSR group are sufficient to re-establish APSR alignment.
The reactor may continue in operation with the APSR misaligned if the limits on power peaking are surveilled within 2 hours to determine if power peaking is still within limits. Also, since any additional movement of the APSRs may result in additional imbalance, Required Action A.1 also requires the power peaking surveillance to be performed again within 2 hours after each APSR movement.
B.1 The unit must be brought to a MODE in which the LCO does not apply if the Required Actions and associated Completion Times cannot be met. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from RTP in an orderly manner and without challenging unit systems. In MODE 3, APSR group alignment limits are not required because the reactor is not generating significant THERMAL POWER and excessive local LHRs cannot occur from APSR misalignment.
SURVEILLANCE REQUIREMENTS SR 3.1.6.1 Verification at a 12 hour Frequency that individual APSR positions are within 6.5% of the group average height limits allows the operator to detect an APSR beginning to deviate from its expected position. In addition, APSR position is continuously available to the operator in the control room so that during actual APSR motion, deviations can immediately be detected. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
: 1. SAR, Section 1.4, GDC 10 and GDC 28.
: 2. 10 CFR 50.46.
: 3. 10 CFR 50.36.
ANO-1                                        B 3.1.6-3                          Amendment No. 215 Rev.
 
Position Indicator Channels B 3.1.7 APPLICABILITY In MODES 1 and 2, OPERABILITY of the position indicator channel is required, since the reactor is, or is capable of, generating THERMAL POWER in these MODES. In MODES 3, 4, 5, and 6, Applicability is not required because the reactor is shut down with the required minimum SDM and is not generating significant THERMAL POWER.
ACTIONS A.1 If the required position indicator channel is inoperable for one or more rods, the position of the CONTROL ROD or APSR is not known with certainty. Therefore, each affected CONTROL ROD or APSR must be declared inoperable, and the limits of LCO 3.1.4 or LCO 3.1.6 apply.
The required Completion Time for declaring the rod(s) inoperable is immediately. Therefore LCO 3.1.4 or LCO 3.1.6 is entered immediately, and the required Completion Times for the appropriate Required Actions in those LCOs apply without delay.
SURVEILLANCE REQUIREMENTS SR 3.1.7.1 A CHANNEL CHECK of the required position indication channel ensures that position indication for each CONTROL ROD and APSR remains OPERABLE and accurate. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. However, this CHANNEL CHECK will be used to detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
When compared to other channels, the agreement criteria between channels is determined by the unit staff. If the channels are within the criteria, it is an indication that the channels are OPERABLE.
The CHANNEL CHECK supplements less formal but more frequent checks of channel OPERABILITY during normal operational use of the displays associated with the LCOs required position indicator channel.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe required Frequency of 12 hours is adequate for verifying that no degradation in system OPERABILITY has occurred.
SR 3.1.7.2 A CHANNEL CALIBRATION of the required position indication channel verifies that the channel responds within the necessary range and accuracy.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months is based on operating experience and consistency with the typical industry refueling cycle.
ANO-1                                            B 3.1.7-3                          Amendment No. 215 Rev.
 
PHYSICS TESTS Exceptions Systems - MODE 1 B 3.1.8 SURVEILLANCE REQUIREMENTS SR 3.1.8.1 Verification that THERMAL POWER is  85% RTP ensures that the required additional thermal margin has been established prior to and during PHYSICS TESTS. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe required Frequency of once per hour allows the operator adequate time to determine any degradation of the established thermal margin during PHYSICS TESTS.
SR 3.1.8.2 Verification that core LHRs are within their limits ensures that core LHR and departure from nucleate boiling ratio will remain within their limits, while one or more of the LCOs that normally control these design limits are out of specification. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe required Frequency of 2 hours allows the operator adequate time for collecting a flux map and for performing the LHR verification, based on operating experience. If SR 3.2.5.1 is not met, PHYSICS TESTS are suspended and LCO 3.2.5 applies. This Frequency is more conservative than the Completion Time for restoration of the individual LCOs that preserve the LHR limits.
This SR is modified by a Note that requires performance only when THERMAL POWER is greater than 20% RTP. This establishes a performance requirement that is consistent with the Applicability of LCO 3.2.5, Power Peaking.
SR 3.1.8.3 Verification that the nuclear overpower trip setpoint is within the limit specified for each PHYSICS TEST ensures that core protection at the reduced power level is established during the PHYSICS TESTS. Performing the verification once within 8 hours prior to the performance of PHYSICS TESTS at each testing plateau allows the operator adequate time for verifying the established trip setpoint before initiating PHYSICS TESTS.
SR 3.1.8.4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:
: a. Reactor Coolant System (RCS) boron concentration;
: b. CONTROL ROD position;
: c. Doppler defect;
: d. Fuel burnup based on gross thermal energy generation;
: e. Samarium concentration;
: f. Xenon concentration; and
: g. Moderator defect.
ANO-1                                          B 3.1.8-5                        Amendment No. 215 Rev.
 
PHYSICS TESTS Exceptions Systems - MODE 1 B 3.1.8 The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 24 hours is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.
REFERENCES
: 1. 10 CFR 50, Appendix B, Section XI.
: 2. 10 CFR 50.59.
: 3. SAR, Section 3A.9.
: 4. SAR, Section 13.3, 13.4 and 13.6.
: 5. SAR, Section 13.4, Table 13-2.
: 6. 10 CFR 50.36.
ANO-1                                      B 3.1.8-5                        Amendment No. 215 Rev.
 
PHYSICS TESTS Exceptions - MODE 2 B 3.1.9 ACTIONS (continued)
B.1 and B.2 If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the unit conditions. Boration will be continued until SDM is within limit. In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied.
Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification. A Completion Time of one hour is provided for the operator to restore compliance with the excepted LCOs.
C.1 If the nuclear overpower trip setpoint is > 5% RTP, then 1 hour is allowed for the operator to restore the nuclear overpower trip setpoint within limits or to complete an orderly suspension of PHYSICS TESTS exceptions. Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable individual LCOs to within specification, in order to ensure that continuity of reactor operation is within initial condition limits. This required Completion Time is consistent with, or more conservative than, those specified for the individual LCOs addressed by PHYSICS TESTS exceptions.
If the nuclear instrumentation high startup rate CONTROL ROD withdrawal inhibit function is inoperable, then 1 hour is allowed for the operator to restore the functions to OPERABLE status or to complete an orderly suspension of PHYSICS TESTS exceptions. Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable individual LCOs to within specification. This required Completion Time is consistent with, or more conservative than, those specified for the individual LCOs addressed by PHYSICS TESTS exceptions.
The nuclear instrumentation high startup rate CONTROL ROD withdrawal inhibit function is not required when the reactor power level is above the operating range of the instrumentation channel. For example, if the reactor power level is above the source range channel operating range, then only the intermediate range high startup rate CONTROL ROD withdrawal inhibit is required to be functional.
SURVEILLANCE REQUIREMENTS SR 3.1.9.1 Verification that THERMAL POWER is  5% RTP ensures that local LHR, DNBR, and RCS pressure limits are not violated and that entry into Actions Condition A is performed promptly.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramHourly verification is adequate for the operator to determine any change in core conditions, such as xenon redistribution occurring after a THERMAL POWER reduction, that could cause THERMAL POWER to exceed the specified limit.
ANO-1                                          B 3.1.9-5                        Amendment No. 215 Rev.
 
PHYSICS TESTS Exceptions - MODE 2 B 3.1.9 SURVEILLANCE REQUIREMENTS (continued)
SR 3.1.9.2 Verification that the nuclear overpower trip setpoint is within the limit specified for PHYSICS TESTS ensures that core protection at the reduced power level is established during PHYSICS TESTS. Performing the verification once within 8 hours prior to the performance of PHYSICS TESTS allows the operator adequate time for verifying the established trip setpoint before initiating PHYSICS TESTS.
SR 3.1.9.3 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:
: a. RCS boron concentration;
: b. CONTROL ROD position;
: c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Samarium concentration;
: f. Xenon concentration;
: g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH);
: h. Moderator defect, when above the POAH; and
: i. Doppler defect, when above the POAH.
Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 24 hours is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.
ANO-1                                        B 3.1.9-6                          Amendment No. 215 Rev.
 
Regulating Rod Insertion Limits B 3.2.1 ACTIONS (continued)
D.2.2 The SDM and ejected rod worth limit can also be restored by reducing the THERMAL POWER to a value allowed by the regulating rod insertion setpoints in the COLR. The required Completion Time of 2 hours is sufficient to allow the operator to complete the power reduction in an orderly manner and without challenging the unit systems. Operation for up to 2 hours in the restricted operation region shown in the COLR is acceptable, based on the low probability of an event occurring simultaneously with the limit out of specification in this relatively short time period. In addition, it precludes long term depletion with abnormal group insertions or configurations and limits the potential for an adverse xenon redistribution.
E.1 If the Required Actions and associated Completion Times of Conditions C or D are not met, then the reactor is placed in MODE 3, in which this LCO does not apply. This Action ensures that the reactor does not continue operating in violation of the peaking limits, the ejected rod worth, the reactivity insertion rate assumed as initial conditions in the accident analyses, or the required minimum SDM assumed in the accident analyses. The required Completion Time of 6 hours is reasonable, based on operating experience regarding the amount of time required to reach MODE 3 from RTP without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.2.1.1 This Surveillance ensures that the sequence and overlap limits are not violated. TheA Surveillance Frequency is controlled under the Surveillance Frequency Control Programof 12 hours is acceptable because little rod motion occurs during this period due to fuel burnup.
Also, the Frequency takes into account other information available in the control room for monitoring the status of the regulating rods.
SR 3.2.1.2 Verification of the regulating rod insertion setpoints as specified in the COLR at a Frequency of 12 hours is sufficient to detect regulating rod banks that may be approaching the group insertion setpoints, because little rod motion due to fuel burnup occurs in 12 hours. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramAlso, the Frequency takes into account other information available in the control room for monitoring the status of the regulating rods.
SR 3.2.1.3 Prior to achieving criticality, an estimated critical position for the CONTROL RODS is determined. Verification that SDM meets the minimum requirements ensures that sufficient SDM capability exists with the CONTROL RODS at the estimated critical position if it is ANO-1                                          B 3.2.1-6                        Amendment No. 215 Rev.
 
APSR Insertion Limits B 3.2.2 ACTIONS (continued)
B.1 If the Required Action and associated Completion Time are not met, the reactor must be placed in MODE 3, in which this LCO does not apply. This action ensures that the fuel does not continue to be depleted in an unintended burnup distribution. The required Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 3 from RTP in an orderly manner and without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.2.2.1 Fuel cycle designs that allow APSR withdrawal near end of cycle (EOC) only permit reinsertion of APSRs at  30% power during plant shutdown for the refueling outage. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramVerification that the APSRs are within their insertion setpoints at a 12 hour Frequency is sufficient to ensure that the APSR insertion setpoints are preserved. The 12 hour Frequency required for performing this verification is sufficient because APSRs are positioned by manual control and are normally moved infrequently. The Frequency takes into account other information available in the control room for monitoring the axial power distribution in the reactor core.
REFERENCES
: 1. SAR, Section 1.4, GDC 10 and GDC 26.
: 2. 10 CFR 50.46.
: 3. SAR, Chapter 14.
: 4. 10 CFR 50.36.
ANO-1                                      B 3.2.2-4                        Amendment No. 215 Rev. 47,
 
AXIAL POWER IMBALANCE Operating Limits B 3.2.3 SURVEILLANCE REQUIREMENTS (continued)
Figure B 3.2.3-1 (Minimum Incore Detector System for AXIAL POWER IMBALANCE Measurement) depicts an example of this configuration. This arrangement is chosen to reduce the uncertainty in the measurement of the AXIAL POWER IMBALANCE by the Minimum Incore Detector System. For example, the requirement for placing one detector of each of the three strings at the core midplane puts three detectors in the central region of the core where the neutron flux tends to be higher. It also helps prevent measuring an AXIAL POWER IMBALANCE that is excessively large when the reactor is operating at low THERMAL POWER levels. The third requirement for placement of detectors (i.e., radial asymmetry) reduces uncertainty by measuring the neutron flux at core locations that are not radially symmetric.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramVerification of the AXIAL POWER IMBALANCE indication every 12 hours ensures that the AXIAL POWER IMBALANCE setpoints are not violated and takes into account other information and alarms available in the control room. This Surveillance Frequency is acceptable because the mechanisms that can cause AXIAL POWER IMBALANCE, such as xenon redistribution or control rod drive mechanism malfunctions that cause slow AXIAL POWER IMBALANCE increases, can be discovered by the operator before the specified limits are violated.
REFERENCES
: 1. 10 CFR 50.46.
: 2. 10 CFR 50.36.
ANO-1                                        B 3.2.3-5                        Amendment No. 215 Rev.
 
QPT B 3.2.4 SURVEILLANCE REQUIREMENTS SR 3.2.4.1 Checking the QPT indication every 7 days ensures that the operator can determine whether the plant computer software and Incore Detector System inputs for monitoring QPT are functioning properly, and takes into account other information and alarms available to the operator in the control room. This procedure allows the QPT mechanisms, such as xenon redistribution, burnup gradients, and CONTROL ROD drive mechanism malfunctions, which can cause slow development of a QPT, to be detected. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramOperating experience has confirmed the acceptability of a Surveillance Frequency of 7 days.
Following restoration of the QPT to within the setpoint, operation at  95% RTP may proceed provided the QPT is determined to remain within the setpoint at the increased THERMAL POWER level. In case QPT exceeds the setpoint for more than 24 hours (Condition A), the potential for xenon redistribution is greater. Therefore, the QPT is monitored for 12 consecutive hourly intervals to determine whether the period of any oscillation due to xenon redistribution causes the QPT to exceed the setpoint again.
REFERENCES
: 1. 10 CFR 50.46
: 2. BAW 10122A, "Normal Operating Controls," Rev. 1, May 1984.
: 3. 10 CFR 50.36 ANO-1                                          B 3.2.4-7                    Amendment No. 215 Rev.
 
RPS Instrumentation B 3.3.1 ACTIONS (continued)
F.1 If Required Action C.1 and Table 3.3.1-1 direct entry into Condition F, the unit must be brought to a MODE in which the specified RPS trip Function is not required to be OPERABLE. To achieve this status, THERMAL POWER must be reduced to < 45% RTP. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach 45% RTP from full power conditions in an orderly manner without challenging unit systems.
G.1 If Required Action C.1 and Table 3.3.1-1 direct entry into Condition G, the unit must be brought to a MODE in which the specified RPS trip Function is not required to be OPERABLE. To achieve this status, THERMAL POWER must be reduced to < 10% RTP. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach 10% RTP from full power conditions in an orderly manner without challenging unit systems.
SURVEILLANCE REQUIREMENTS The SRs for each RPS Function are identified by the SRs column of Table 3.3.1-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION testing.
The SRs are modified by a Note which directs the reader to Table 3.3.1-1 to determine the correct SRs to perform for each RPS Function.
SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours provides reasonable assurance of prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to the same 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 two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff based on a combination of factors including channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Off scale low current loop channels are, where practical, verified to be reading at the bottom of the range and not failed downscale.
ANO-1                                          B 3.3.1-17                      Amendment No. 215 Rev.
 
RPS Instrumentation B 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.1 (continued)
The agreement criteria includes an expectation of one decade of overlap when transitioning between neutron flux instrumentation. For example, during a power increase near the top of the scale for the intermediate range monitors, a power range monitor reading is expected with at least one decade overlap. Without such an overlap, the power range monitors are considered inoperable unless it is clear that an intermediate range monitor inoperability is responsible for the lack of the expected overlap.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal but more frequent checks of channel OPERABILITY during normal operational use of the displays associated with the LCO's required channels.
For Functions that trip on a combination of several measurements, such as the Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE Function, the CHANNEL CHECK must be performed on each input.
SR 3.3.1.2 This SR is the performance of a heat balance calibration for the power range channels periodicallyevery 96 hours and once within 24 hours after a THERMAL POWER change of 10% RTP in one direction, when reactor power is  20% RTP. The heat balance calibration consists of a comparison of the results of the calorimetric with the power range channel output.
The outputs of the power range channels are calibrated to the calorimetric. Note 1 to the SR states if the absolute difference between the calorimetric and the Nuclear Instrumentation System (NIS) channel is> 2% RTP, the NIS channel is not declared inoperable but must be adjusted. If the NIS channel cannot be properly adjusted, the channel is declared inoperable.
Note 2 clarifies that this Surveillance is required only if reactor power is  20% RTP and that 24 hours is allowed for performing the first Surveillance after reaching 20% RTP.
Two calorimetric calculations are routinely performed. One relies upon primary system parameters and the other relies upon secondary system parameters. The primary calorimetric is generally less accurate than the secondary calorimetric at higher power levels and more accurate at lower power levels. For comparison to the nuclear instrumentation, between 0 and 15% power, only the primary calorimetric (heat balance) is considered. From 15 to 100% power the calorimetric is weighted linearly with only the secondary heat balance being considered at 100% power.
The power range channel's output shall be adjusted consistent with the calorimetric results if the absolute difference between the calorimetric and the power range channel's output is > 2%
RTP. The value of 2% is adequate because this value is assumed in the safety analyses of SAR, Chapter 14 (Ref. 2). These checks and, if necessary, the adjustment of the power range channels ensure that channel accuracy is maintained within the analyzed error margins.
ANO-1                                        B 3.3.1-18                        Amendment No. 215 Rev.
 
RPS Instrumentation B 3.3.1 The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 96 hour Frequency is adequate, based on unit operating experience, which demonstrates the change in the difference between the power range indication and the calorimetric results rarely exceeds 2% in any 96 hour period. Furthermore, the control room operators monitor redundant indications and alarms to detect deviations in channel outputs.
ANO-1                                      B 3.3.1-18                      Amendment No. 215 Rev.
 
RPS Instrumentation B 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.3 A comparison of power range nuclear instrumentation channels against incore detectors shall be performed periodicallyat a 31 day Frequency when reactor power is  20% RTP. The SR is modified by two Notes. Note 2 clarifies that 24 hours is allowed for performing the first Surveillance after reaching 20% RTP. Note 1 states if the absolute difference between the power range and incore AXIAL POWER IMBALANCE measurements is  2% RTP, the power range channel is not inoperable, but an adjustment of the measured imbalance to agree with the incore measurements is necessary. If the power range channel cannot be properly recalibrated, the channel is declared inoperable. The calculation of the Allowable Value envelope assumes a difference in out of core to incore AXIAL POWER IMBALANCE measurements of 2.5%.
Additional inaccuracies beyond those that are measured are also included in the setpoint envelope calculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is adequate, considering that long term drift of the excore linear amplifiers is small and burnup of the detectors is slow. Also, the excore readings are a strong function of the power produced in the peripheral fuel bundles, and do not represent an integrated reading across the core. The slow changes in neutron flux during the fuel cycle can also be detected at this interval.
SR 3.3.1.4 A CHANNEL FUNCTIONAL TEST is performed to ensure that the entire channel will perform the intended function. Setpoints must be found within the Allowable Values specified in Table 3.3.1-1. Any setpoint adjustment shall be consistent with the assumptions of the current setpoint analysis.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 31 days is based on operating experience, which has demonstrated through high reliability of the instrumentation, that failure of more than one channel, of a given Function, in any 31 day interval is rare. Testing in accordance with this SR is normally performed on a rotational basis, with one channel being tested each week. Testing one channel each week reduces the probability of an undetected failure existing within the system and minimizes the likelihood of the same systematic test errors being introduced into each redundant channel. The automatic bypass removal feature is verified for the turbine oil pressure trip and the main feedwater pump oil pressure trip functions during the CHANNEL FUNCTIONAL TEST.
SR 3.3.1.5 A Note to the Surveillance indicates that neutron detectors are excluded from CHANNEL CALIBRATION. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.
A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The 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 drift to ensure that the instrument channel remains operational between ANO-1                                        B 3.3.1-19                        Amendment No. 215 Rev.
 
RPS Instrumentation B 3.3.1 successive tests. CHANNEL CALIBRATION shall find that instrument errors are within the assumptions of the setpoint analysis. CHANNEL CALIBRATION must be performed consistent with the assumptions of the setpoint analysis. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature (RTD) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is justified by the assumption of at least an 18 month calibration interval in the determination of the allowable magnitude of equipment drift in the setpoint analysis.
ANO-1                                      B 3.3.1-19                      Amendment No. 215 Rev.
 
RPS - RTM B 3.3.3 ACTIONS (continued)
A.1.1, A.1.2, and A.2 (continued)
To ensure the trip signal is registered in the other channels, Required Action A.2 requires that the inoperable RTM be removed from the cabinet. This action causes the electrical interlocks to indicate a tripped channel in the remaining three RTMs. Operation in this condition is allowed indefinitely because the actions put the RPS into a one-out-of-three configuration. The 1 hour Completion Time is sufficient time to perform the Required Actions.
B.1, B.2.1, and B.2.2 Condition B applies if two or more RTMs are inoperable in MODE 1, 2, or 3, or if the Required Actions and associated Completion Time of Condition A are not met in MODE 1, 2, or 3. In this case, the unit must be placed in a MODE in which the LCO does not apply. This is done by placing the unit in at least MODE 3 with all CRD trip breakers open or with power to all CRD trip breakers removed within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.
C.1 and C.2 Condition C applies if two or more RTMs are inoperable in MODE 4 or 5, or if the Required Actions and associated Completion Times are not met in MODE 4 or 5. In this case, the unit must be placed in a MODE in which the LCO does not apply. This is done by opening all CRD trip breakers or removing power from all CRD trip breakers. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to open all CRD trip breakers or remove all power to the CRD System without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.3.3.1 The SRs include performance of a CHANNEL FUNCTIONAL TEST every 92 days. This test shall verify the OPERABILITY of the RTM and its ability to receive and properly respond to channel trip and reactor trip signals.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 92 days is based on operating experience, which has demonstrated through high reliability of the instrumentation, that failure of more than one RTM in any 92 day interval is rare (Ref. 3).
Testing in accordance with this SR is normally performed on a rotational basis, with one RTM being tested each 23 days. Testing one RTM each 23 days reduces the probability of an undetected failure existing within the system and minimizes the likelihood of the same systematic test errors being introduced into each redundant RTM.
ANO-1                                        B 3.3.3-3                        Amendment No. 215 Rev.
 
CRD Trip Devices B 3.3.4 ACTIONS (continued)
E.1 and E.2 If the Required Actions and associated Completion Times of Condition A, B, or C are not met in MODE 4 or 5, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, all CRD trip breakers must be opened or power to all CRD trip breakers removed within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to open all CRD trip breakers or remove power from all CRD trip breakers without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.3.4.1 SR 3.3.4.1 is to perform a CHANNEL FUNCTIONAL TEST every 92 days. This test verifies the OPERABILITY of the trip devices by actuation of the end devices. Also, this test independently verifies the undervoltage and shunt trip mechanisms of the trip breakers. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program The Frequency of 92 days is based on operating experience, which has demonstrated that failure of more than one channel of a given function in any 92 day interval is a rare event (Ref. 3).
Testing in accordance with this SR is normally performed on a rotational basis with one channel being tested each 23 days. Testing one channel each 23 days reduces the probability of an undetected failure existing within the system and minimizes the likelihood of the same systematic test errors being introduced into each redundant trip device.
REFERENCES
: 1. SAR, Chapter 7.
: 2. 10 CFR 50.36.
: 3. BAW-10167A, Justification for Increasing the Reactor Trip System On-Line Test Intervals, Supplement 3, Justification for Increasing the Trip Device Test Interval, February 1998.
ANO-1                                        B 3.3.4-5                        Amendment No. 215 Rev.
 
ESAS Instrumentation B 3.3.5 SR 3.3.5.1 Performance of the CHANNEL CHECK every 12 hours provides reasonable assurance for prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between CHANNEL CALIBRATIONs.
Agreement criteria are determined by the unit staff, based on a combination of factors including channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12-hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but potentially more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO's required channels.
SR 3.3.5.2 A CHANNEL FUNCTIONAL TEST is performed on each required ESAS analog instrument channel to ensure the entire channel will perform the intended functions. Any setpoint adjustment shall be consistent with the assumptions of the setpoint calculations.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 31 days is based on unit operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given function in any 31-day interval is a rare event. The RCS low pressure automatic bypass removal feature is verified during its CHANNEL FUNCTIONAL TEST.
SR 3.3.5.3 CHANNEL CALIBRATION is a complete check of the analog instrument channel, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the analog instrument channel remains OPERABLE between successive tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the setpoint calculations. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint calculations.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThis Frequency is justified by the assumption of at least an 18-month calibration interval to determine the magnitude of equipment drift in the setpoint calculations.
ANO-1                                          B 3.3.5-10                      Amendment No. 215 Rev. 50,
 
ESAS Manual Initiation B 3.3.6 SURVEILLANCE REQUIREMENTS SR 3.3.6.1 This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the ESAS manual initiation. This test verifies that the initiating circuitry is OPERABLE and will actuate the digital actuation logic channels. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency is demonstrated to be sufficient, based on operating experience, which shows these components usually pass the Surveillance when performed on the 18-month Frequency.
REFERENCES
: 1. 10 CFR 50.36.
: 2. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
ANO-1                                            B 3.3.6-4                      Amendment No. 215 Rev. 50,
 
ESAS Actuation Logic B 3.3.7 SURVEILLANCE REQUIREMENTS SR 3.3.7.1 SR 3.3.7.1 is the performance of a CHANNEL FUNCTIONAL TEST on a 31 day Frequency.
The test demonstrates that each digital actuation logic channel successfully performs the two-out-of-three logic combinations every 31 days. The test simulates the required one-out-of-three inputs to the logic circuit and verifies the successful operation of the digital actuation logic. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates the rarity of more than one channel failing within the same 31 day interval. The CHANNEL FUNCTIONAL TEST performed for the Reactor Building Spray System Logic Channels shall include testing of the associated spray pump, spray valves, and chemical additive valve logic channels.
REFERENCES
: 1. 10 CFR 50.46.
: 2. BAW-10192PA, BWNT Loss-of-Coolant Accident Evaluation Model for OTSG Plants, Volumes I and II, June 1998.
: 3. 10 CFR 50.36.
ANO-1                                          B 3.3.7-4                          Amendment No. 215 Rev. 1,
 
DG LOPS B 3.3.8 B.1 Condition B applies if the Required Action and associated Completion Time of Condition A are not met.
Required Action B.1 ensures that Required Actions for affected diesel generator inoperabilities are initiated. Depending on the DG(s) affected, the appropriate Actions specified in LCO 3.8.1, "AC Sources - Operating," are required immediately.
SURVEILLANCE REQUIREMENTS SR 3.3.8.1 Performance of the CHANNEL CHECK once every 7 days provides reasonable assurance for prompt identification of a gross failure of instrumentation. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates channel failure is rare. Since the probability of random failure in any 7 day period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of this instrumentation.
SR 3.3.8.2 The Note allows channel bypass for testing of the loss of voltage Function without entering the associated Conditions and Required Actions, although during this time period it cannot actuate a diesel start. This allowance is based on the assumption that 4 hours is the average time required to perform channel Surveillance. The 4 hour testing allowance does not significantly reduce the probability that the DG will start when necessary. It is not acceptable to remove channels from service for more than 4 hours to perform required Surveillance testing without declaring the channel inoperable.
A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The setpoints and the response to a loss of voltage and a degraded voltage test shall include a single point verification that the trip occurs within the required delay time. CHANNEL CALIBRATION shall verify that setpoints are within the required ranges. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on the reliability of the components, on operating experience which demonstrates channel failure is rare, and on consistency with the typical industry refueling cycle, and is justified by the assumption of at least an 18 month calibration interval in the determination of equipment drift.
ANO-1                                            B 3.3.8-4                      Amendment No. 215 Rev. 56,
 
Source Range Neutron Flux B 3.3.9 SURVEILLANCE REQUIREMENTS SR 3.3.9.1 Performance of the CHANNEL CHECK once every 12 hours provides reasonable assurance of prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to the same parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of factors including channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction.
The agreement criteria includes an expectation of one decade of overlap when transitioning between neutron flux instrumentation. For example, during a power reduction near the bottom of the scale for the intermediate range monitors, a source range monitor reading is expected with at least one decade overlap. Without such an overlap, the source range monitors are considered inoperable unless it is clear that an intermediate range monitor inoperability is responsible for the lack of the expected overlap.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO's required channel. When operating with only one channel OPERABLE, CHANNEL CHECK is still required. However, in this condition, a redundant source range may not be available for comparison. CHANNEL CHECK may still be performed via comparison with intermediate range detectors, if available, and verification that the OPERABLE source range channel is energized and indicating a value consistent with current unit status.
SR 3.3.9.2 For a source range neutron flux channel, CHANNEL CALIBRATION is a complete check and readjustment of the channel from the preamplifier input to the indicator. This test verifies the channel responds to measured parameters within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel at a setpoint which accounts for instrument drift to ensure that the instrument channel remains operational between successive tests.
ANO-1                                            B 3.3.9-3                      Amendment No. 215 Rev.
 
Source Range Neutron Flux B 3.3.9 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.9.2 (continued)
The SR is modified by a Note excluding neutron detectors from CHANNEL CALIBRATION. It is not necessary to test the detectors because generating a meaningful test signal is difficult, and there is no adjustment that can be made to the detectors. Furthermore, adjustment of the detectors is unnecessary because they are passive devices with minimal drift. Finally, the detectors are of simple construction, and any failures in the detectors will be apparent as change in channel output.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months is based on demonstrated instrument CHANNEL CALIBRATION reliability over an 18 month interval, such that the instrument is not adversely affected by drift.
REFERENCES
: 1. 10 CFR 50.36.
ANO-1                                      B 3.3.9-4                          Amendment No. 215 Rev.
 
Intermediate Range Neutron Flux B 3.3.10 APPLICABILITY (continued)
The intermediate range instrumentation is designed to detect power changes when the power range and source range instrumentation cannot provide reliable indications, e.g., during initial criticality and power escalation. Since those conditions can exist in, or propagate from, all of these MODES, the intermediate range instrumentation must be OPERABLE.
ACTIONS A.1 and A.2 With the required intermediate range neutron flux channel inoperable when THERMAL POWER is  5% RTP, the operators must place the reactor in the next lowest condition for which the intermediate range instrumentation is not required. This involves providing power level indication on the source range instrumentation by immediately suspending operations involving positive reactivity changes and, within 1 hour, placing the reactor in the tripped condition with the CRD trip breakers open. RCS temperature changes are permitted provided the effects of such changes are accounted for in the SDM calculations. The Completion Times are based on unit operating experience and allow the operators sufficient time to manually insert the CONTROL RODS prior to opening the CRD breakers.
SURVEILLANCE REQUIREMENTS SR 3.3.10.1 Performance of the CHANNEL CHECK once every 12 hours provides reasonable assurance of prompt identification that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to the same parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff based on a combination of factors including channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. Off scale low current loop channels are verified, where practical to be reading at the bottom of the range and not failed low.
The agreement criteria includes an expectation of one decade of overlap when transitioning between neutron flux instrumentation. For example, during a power increase near the top of the scale for the source range monitors, an intermediate range monitor reading is expected with at least one decade overlap. Without such an overlap, the intermediate range monitors are considered inoperable unless it is clear that a source range monitor inoperability is responsible for the lack of the expected overlap.
ANO-1                                          B 3.3.10-2                        Amendment No. 215 Rev.
 
Intermediate Range Neutron Flux B 3.3.10 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.10.1 (continued)
Further, during a power reduction near the bottom of the scale for the power range monitors, an intermediate range monitor reading is expected with at least one decade overlap. Without such an overlap, the intermediate range monitors are considered inoperable unless it is clear that a power range monitor inoperability is responsible for the lack of the expected overlap.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency, about once every shift, is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12-hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO's required channel.
When operating with only one channel OPERABLE, CHANNEL CHECK is still required.
However, in this condition, a redundant intermediate range is not available for comparison.
CHANNEL CHECK may still be performed via comparison with power or source range detectors, if available, and verification that the OPERABLE intermediate range channel is energized and indicates a value consistent with current unit status.
SR 3.3.10.2 A CHANNEL FUNCTIONAL TEST, of the required intermediate range instrument channel, verifies proper operation of the channel each 31 days. Monthly testing provides reasonable assurance that the instrument channel will function, if required, to provide indication during MODE 2 and during unanticipated reactivity excursions from MODES 3, 4, or 5. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.10.3 For intermediate range neutron flux channels, CHANNEL CALIBRATION is a complete check and readjustment of the channels, from the preamplifier input to the indicators. This test verifies the channel responds to a measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel at a setpoint which accounts for instrument drift to ensure that the instrument channel remains operational between successive tests.
The SR is modified by a Note excluding neutron detectors from CHANNEL CALIBRATION. It is not necessary to test the detectors because generating a meaningful test signal is difficult. In addition, the detectors are of simple construction, and any failures in the detectors will be apparent as a change in channel output. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience and consistency with the typical industry refueling cycle and is justified by demonstrated instrument reliability over an 18-month interval such that the instrument is not adversely affected by drift.
ANO-1                                          B 3.3.10-3                      Amendment No. 215 Rev.
 
EFIC Instrumentation B 3.3.11 SURVEILLANCE REQUIREMENTS A Note indicates that the SRs for each EFIC instrumentation Function are identified in the SRs column of Table 3.3.11-1. Individual EFIC subgroup relays must also be tested, one at a time, to verify the individual EFIC components will actuate when required. Some components cannot be tested at power since their actuation might lead to unit trip or equipment damage. These are specifically identified and must be tested when shut down. The various SRs account for individual functional differences and for test frequencies applicable specifically to the Functions listed in Table 3.3.11-1. The operational bypasses associated with each EFIC instrumentation channel are also subject to these SRs to ensure OPERABILITY of the EFIC instrumentation channel.
SR 3.3.11.1 Performance of the CHANNEL CHECK once every 12 hours provides reasonable assurance for prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of factors including channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Off scale low current loop channels are verified, where practical, to be reading at the bottom of the range and not failed downscale.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels.
SR 3.3.11.2 A CHANNEL FUNCTIONAL TEST verifies the function of the automatic bypass removal feature, required trip, interlock, and alarm functions of the channel. Setpoints for trip functions must be found within the Allowable Value. (Note that the values for the bypass removal functions are identified in the Applicable MODES or Other Specified Condition column of Table 3.3.11-1 as limits on applicability for the trip Functions.) Any setpoint adjustment shall be consistent with the assumptions of the current setpoint analysis.
ANO-1                                          B 3.3.11-12                    Amendment No. 215 Rev. 12,
 
EFIC Instrumentation B 3.3.11 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.11.2 (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 31 days is based on unit operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given function in any 31 day interval is a rare event.
This SR is modified by two notes. For the SG Level - Low function, if the as-found trip setpoint is found to be non-conservative with respect to the Allowable Value specified in TSs, the channel is declared inoperable and the associated TS action statement must be followed. If the as-found trip setpoint is found to be conservative with respect to the Allowable Value and outside the as-found predefined acceptance criteria band of +/- 1.08 inches from the previous as-left value, but is determined to be functioning as required and can be reset to a value equal to the Limiting Trip Setpoint or a value more conservative than the Limiting Trip Setpoint, then the channel may be considered to be operable. If it cannot be determined that the instrument channel is functioning as required, the channel is declared inoperable and the associated TS actions must be followed. If the as-found trip setpoint is outside the as-found predefined acceptance criteria band, the condition must be entered into the corrective action program for further evaluation. The notes for the Channel Functional Test do not apply to the verification of the time delay.
SR 3.3.11.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor.
The test verifies the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channels adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive tests.
CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the setpoint analysis. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint analysis. The notes contained in SR 3.3.11.2 are also applicable to the CHANNEL CALIBRATION.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on the assumption of at least an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
REFERENCES
: 1. 10 CFR 50.62.
: 2. SAR, Chapter 7.
: 3. SAR, Chapter 14.
: 4. Instrument Loop Error Analysis and Setpoint Methodology Manual, Design Guide, IDG-001.
ANO-1                                        B 3.3.11-13                      Amendment No. 215 Rev. 12,13,
 
EFIC Manual Initiation B 3.3.12 ACTIONS (continued)
C.1 With one or both required manual initiation switches of one or more EFIC Function(s) inoperable in both actuation trains, one actuation train for each Function must be restored to OPERABLE status within 1 hour. With the train restored, the second train must be placed in the appropriate condition within 72 hours per Required Action A.1 or B.1, as applicable. Compliance with these actions ensures the single-failure criterion is met. The Completion Time allotted to restore the train allows the operator to take all the appropriate actions for the failed train and still ensures that the risk involved in operating with the failed train is acceptable.
D.1 and D.2 If the Required Action and the associated Completion Time is not met for any EFW Initiation Function, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODES from full power conditions in an orderly manner and without challenging unit systems.
E.1, E.2.1, and E.2.2 If the Required Actions and associated Completion Times are not met for the Main Steam Line Isolation Function, the unit must be placed in a MODE or condition in which the requirement does not apply. This is initiated by placing the unit in MODE 3 within 6 hours and, either reducing SG pressure to less than 750 psig, or closing and deactivating all associated valves, i.e., the valves which EFIC would close if it were to actuate while OPERABLE. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.3.12.1 This SR requires the performance of a CHANNEL FUNCTIONAL TEST to ensure that the trains can perform their intended functions. However, for Main Steam Line Isolation and EFW Initiation, the test need not include actuation of the end device. This is due to the risk of a unit transient caused by the closure of valves associated with Main Steam Line Isolation or EFW Initiation during testing at power. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 31 days is based on operating experience with regard to channel OPERABILITY that demonstrates the rarity of more than one train failing within the same 31 day interval.
REFERENCES
: 1. IEEE-279-1971, April 1972.
ANO-1                                        B 3.3.12-3                        Amendment No. 215 Rev.
 
EFIC Logic B 3.3.13 ACTIONS (continued)
A.1 (continued)
Condition A can be thought of as equivalent to failure of a single train of a two train safety system (e.g., the safety function can be accomplished, but a single failure cannot be taken).
Thus, the Completion Time of 72 hours has been chosen to be consistent with Completion Times for restoring one inoperable ESF System train.
The EFIC System has not been analyzed for failure of both trains of the same Function.
Consequently, any combination of failures in both trains A and B is not covered by Condition A and must be addressed by entry into LCO 3.0.3.
B.1 and B.2 If Required Action A.1 and its associated Completion Time is not met for the EFW Initiation Function, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODES from full power conditions in an orderly manner and without challenging unit systems.
C.1, C.2.1, and C.2.2 If the Required Actions and associated Completion Times are not met for the Main Steam Line Isolation Function, the unit must be placed in a MODE or condition in which the requirement does not apply. This is initiated by placing the unit in MODE 3 within 6 hours and, either reducing SG pressure to less than 750 psig, or closing and deactivating all associated valves, i.e., the valves which EFIC would close if it were to actuate while OPERABLE. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.3.13.1 This SR requires the performance of a CHANNEL FUNCTIONAL TEST to ensure that the trains can perform their intended functions. This test verifies Main Steam Line Isolation and EFW Initiation automatic actuation logics are functional. This test simulates the required inputs to the logic circuit and verifies successful operation of the automatic actuation logic. The test need not include actuation of the end device. This is due to the risk of a unit transient caused by the closure of valves associated with Main Steam Line Isolation or actuation of EFW during testing at power. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 31 days is based on operating experience with regard to channel OPERABILITY, which has demonstrated the rarity of more than one channel failing within the same 31 day interval.
ANO-1                                        B 3.3.13-4                        Amendment No. 215 Rev.
 
EFIC Vector Logic B 3.3.14 SURVEILLANCE REQUIREMENTS SR 3.3.14.1 SR 3.3.14.1 is the performance of a CHANNEL FUNCTIONAL TEST every 31 days. This test demonstrates that the EFIC vector logic performs its function as desired. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience with respect to channel OPERABILITY that demonstrates the rarity of more than one channel failing within the same 31 day interval.
REFERENCES None.
ANO-1                                        B 3.3.14-4                    Amendment No. 215 Rev.
 
PAM Instrumentation B 3.3.15 ACTIONS (continued)
F.1 (continued)
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.
The Special Report is to be submitted in accordance with 10 CFR 50.4 within 30 days of entering Condition F.
Both the RCS Hot Leg Level and the Reactor Vessel Level are methods of monitoring for inadequate core cooling.
The alternate means of monitoring the Reactor Building Area Radiation (High Range) consist of a combination of installed area radiation monitors and portable instrumentation.
The Completion Time of Immediately for Required Action F.1 identifies the start of the clock for submittal of the Special Report.
SURVEILLANCE REQUIREMENTS As noted at the beginning of the SRs, the SRs apply to each PAM instrumentation Function in Table 3.3.15-1.
SR 3.3.15.1 Performance of the CHANNEL CHECK once every 31 days for each required instrumentation channel that is normally energized provides reasonable assurance for prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel with a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared with similar unit instruments located throughout the unit. For the reactor building hi-range radiation monitor, the CHANNEL CHECK should also note the detector's response to the keep alive source.
Agreement criteria are determined by the unit 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.
If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Offscale low current loop channels are, where practical, verified to be reading at the bottom of the range and not failed downscale.
ANO-1                                        B 3.3.15-10                        Amendment No. 215 Rev. 5,22,
 
PAM Instrumentation B 3.3.15 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.15.1 (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on unit operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal but more frequent checks of channels during normal operational use of the displays associated with this LCO's required channels.
SR 3.3.15.2 A CHANNEL CALIBRATION is performed every 18 months. CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. This test verifies the channel responds to measured parameters within the necessary range and accuracy.
The SR is modified by a Note excluding neutron detectors from CHANNEL CALIBRATION. It is not necessary to test the detectors because generating a meaningful test signal is difficult, and there is no adjustment that can be made to the detectors. Furthermore, adjustment of the detectors is unnecessary because they are passive devices, with minimal drift. Finally, the detectors are of simple construction, and any failures in the detectors will be apparent as change in channel output.
For the Reactor Building Area Radiation instrumentation, a CHANNEL CALIBRATION may 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 a gamma source.
Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detector (RTD) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience and consistency with the typical industry refueling cycle and is justified by the assumption of at least an 18 month calibration interval in the determination of the magnitude of equipment drift.
REFERENCES
: 1. SAR, Table 7-11A.
: 2. Regulatory Guide 1.97.
: 3. NUREG-0737, 1979.
ANO-1                                        B 3.3.15-11                      Amendment No. 215 Rev. 5,22,
 
Control Room Isolation - High Radiation B 3.3.16 ACTIONS (continued)
C.1 and C.2 If the CREVS cannot be placed into the emergency recirculation mode while in MODE 1, 2, 3, or 4, actions must be taken to minimize the chances of an accident that could lead to radiation releases. The unit must be placed in at least MODE 3 within 6 hours, with a subsequent cooldown to MODE 5 within 36 hours. This places the reactor in a low energy state that allows greater time for operator action if habitation of the control room is precluded. 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.
D.1 and D.2 Required Action D.1 is the same as discussed earlier for Condition A, except for Completion Time. If the CREVS cannot be placed into recirculation mode while moving irradiated fuel assemblies, then Required Action D.2 suspends actions that could lead to an accident that could release radioactivity resulting from a fuel handling accident.
Required Action D.2 places the irradiated fuel in a safe and stable configuration in which it is less likely to experience an accident that could result in a release of radioactivity. The irradiated fuel must be maintained in these conditions until the automatic isolation capability is returned to operation or when manual action places one train of the CREVS into the emergency recirculation mode. The Completion Time of "Immediately" is consistent with the urgency of the situation and accounts for the high radiation function, which provides the only automatic Control Room Isolation function capable of responding to radiation release due to a fuel handling accident. The Completion Time does not preclude placing any fuel assembly into a safe position before ceasing any such movement.
Note that in certain circumstances, such as fuel handling in the fuel handling area during power operation, both Condition A or B and Condition D may apply in the event of channel failure(s).
SURVEILLANCE REQUIREMENTS SR 3.3.16.1 Performance of a CHANNEL CHECK for the Control Room Isolation - High Radiation actuation instrumentation once every 12 hours provides reasonable assurance for prompt identification of a gross failure of instrumentation. Performance of the CHANNEL CHECK helps ensure that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Acceptance criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on operating experience that demonstrates channel failure is rare.
ANO-1                                            B 3.3.16-3                    Amendment No. 215 Rev.
 
Control Room Isolation - High Radiation B 3.3.16 SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.16.2 A Note allows a channel to be inoperable for up to 3 hours for surveillance testing without entering the associated Conditions and Required Actions, although during this time period it cannot actuate a control room isolation. This is based on the average time required to perform channel surveillance. It is not acceptable to remove channels from service for more than 3 hours to perform required surveillance testing without declaring the channel inoperable.
SR 3.3.16.2 is the performance of a CHANNEL FUNCTIONAL TEST once every 31 days to ensure that the channels can perform their intended functions. This test verifies the capability of the instrumentation to provide the automatic Control Room Isolation. Any setpoint adjustment shall be consistent with the setpoint requirements.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is based on operating experience which indicates that the instrumentation usually passes the CHANNEL FUNCTIONAL TEST when performed on a monthly basis.
SR 3.3.16.3 This SR requires the performance of a CHANNEL CALIBRATION to ensure that the instrument channel remains operational with the correct setpoint.
CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATION must be performed consistent with the setpoint requirements.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on the assumption of at least an 18 month calibration interval in the determination of the magnitude of equipment drift and is consistent with the typical refueling cycle.
REFERENCES
: 1. ANO-2 SAR, Section 6.4.
: 2. 10 CFR 50.36.
ANO-1                                        B 3.3.16-4                      Amendment No. 215 Rev.
 
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 APPLICABILITY (continued)
The Note indicates the limit on RCS pressure may be exceeded during short term operational pressure transients resulting from a THERMAL POWER change > 5% RTP per minute. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, for transients initiated from power levels less than the Allowable Thermal Power, increased DNBR margin exists to offset the temporary pressure variations.
ACTIONS A.1 Loop pressure and hot leg coolant temperature are controllable and measurable parameters.
With one or both of these parameters not within the LCO limits, action must be taken to restore the parameters. RCS flow rate is not a controllable parameter and is not expected to vary during steady state operation. However, if the flow rate is below the LCO limit, the parameter must be restored to within limits or power must be reduced as required in Required Action B.1, to eliminate the potential for violation of the minimum DNBR limit.
The 2-hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, determine the cause for the off normal condition, and restore the readings within limits. The Completion Time is based on plant operating experience.
B.1 If the Required Action and associated Completion Time are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis assumptions.
The 6-hour Completion Time is reasonable, based on operating experience, to reach MODE 2 from full power conditions in an orderly manner and without challenging safety systems.
SURVEILLANCE REQUIREMENTS SR 3.4.1.1 The RCS pressure value specified is dependent on the number of pumps in operation and has been adjusted to account for the pressure difference between the core exit and the measurement location. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify operation is within safety analysis assumptions.
A Note has been added to indicate the pressure limits are to be applied to the loop with two pumps in operation for the three pump operating condition.
ANO-1                                          B 3.4.1-3                      Amendment No. 215 Rev.
 
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.1.2 The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify that operation is within safety analysis assumptions.
A Note has been added to indicate the temperature limits are to be applied to the loop with two pumps in operation for the three pump operating condition.
SR 3.4.1.3 The 12 hour Surveillance Frequency for RCS total flow rate is performed using the available flow indications. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify that operation is within safety analysis assumptions.
SR 3.4.1.4 Measurement of RCS total flow rate once every 18 months allows the installed RCS flow instrumentation to be calibrated and verifies that the actual RCS flow is greater than or equal to the minimum required RCS flow rate specified in the COLR.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months reflects the importance of verifying flow after a refueling outage when the core has been altered or RCS flow characteristics may have been modified, which may have caused change of flow.
The Surveillance is modified by a Note that indicates the SR does not need to be performed until stable thermal conditions are established at higher power levels (i.e.,  90% RTP). The Note provides for measurement of the flow rate at normal operating conditions at power in MODE 1. The Surveillance may be performed at low power or in MODE 2 or below. However, at low or zero power conditions, the indications are less accurate and significant penalties for uncertainties may be necessary. Performance of the calorimetric heat balance at a high power level and normal operating conditions provides for the most accurate flow verification.
REFERENCES
: 1. SAR, Chapter 14.
: 2. SAR, Section 3A.6.
: 3. BAW-10179P-A, Rev. 6, August 2005.
: 4. 10 CFR 50.36.
ANO-1                                        B 3.4.1-4                        Amendment No. 215 Rev. 9,
 
RCS Minimum Temperature for Criticality B 3.4.2 APPLICABILITY The reactor has been designed and analyzed to be critical in MODES 1 and 2 only with Tavg 525&deg;F. Criticality is not permitted in any other MODE. Therefore, this LCO is applicable in MODE 1 and MODE 2.
ACTIONS A.1 With Tavg below 525 &deg;F, 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 in 30 minutes. Rapid reactor shutdown can be readily and practically achieved in a 30-minute period. The Completion Time reflects the ability to perform this Action and maintain the plant within the analyzed range. If Tavg can be restored within the 30 minute time period, shutdown is not required.
SURVEILLANCE REQUIREMENTS SR 3.4.2.1 RCS average temperature is required to be verified periodically at or above 525 &deg;F every 12 hours. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe SR to verify RCS average temperature every 12 hours takes into account indications that are continuously available to the operator in the control room and is consistent with other routine surveillances which are typically performed once per shift. In addition, Operators are trained to be sensitive to RCS temperature during approach to criticality and will ensure that the minimum temperature for criticality is met as criticality is approached.
RCS Tavg is normally calculated as the average of the unit Thot (hot temperature average of loops A and B) and the unit Tcold (cold temperature average of loops A and B). During operation with 3 RCPs in operation, Tavg is calculated as the average of the loop Thot and loop Tcold in the loop that has 2 RCPs running.
REFERENCES
: 1. SAR, Chapter 14.
: 2. 10 CFR 50.36.
ANO-1                                        B 3.4.2-2                        Amendment No. 215 Rev.
 
RCS P/T Limits B 3.4.3 ACTIONS (continued)
C.1 and C.2 (continued)
If the required evaluation for continued operation cannot be accomplished within 72 hours, or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Actions C.1 and C.2. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions.
However, if the favorable evaluation is accomplished while reducing pressure and temperature conditions, a return to power operation may be initiated without completing these Required Actions.
Pressure and temperature are reduced by bringing the unit to MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging unit systems.
D.1 and D.2 Actions must be initiated immediately to correct operation outside of the P/T limits at times other than MODE 1, 2, 3, or 4, so that the RCPB is returned to a condition that has been verified acceptable by stress analysis.
The immediate Completion Time reflects the urgency of initiating action to restore the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished promptly in a controlled manner.
In addition to restoring operation to within limits, an evaluation is required to verify that the RCPB integrity remains acceptable. The evaluation must be completed prior to entry into MODE 4. Several methods may be used, including comparison with pre-analyzed transients in the stress analysis, or inspection of the components.
ASME Code, Section XI, Appendix E (Ref. 10), may also be used to support the evaluation.
However, its use is restricted to evaluation of the vessel beltline.
Condition D is modified by a Note requiring Required Action D.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone, per Required Action D.1, is insufficient because higher than analyzed stresses may have occurred and may have affected RCPB integrity.
SURVEILLANCE REQUIREMENTS SR 3.4.3.1, SR 3.4.3.2, SR 3.4.3.3, and SR 3.4.3.4 Verification that operation is within the limits of the appropriate figure is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.
The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.
ANO-1                                          B 3.4.3-6                        Amendment No. 215 Rev. 51,
 
RCS P/T Limits B 3.4.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.3.1, SR 3.4.3.2, SR 3.4.3.3, and SR 3.4.3.4 (continued)
This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.
Surveillance for heatup, cooldown, or inservice hydrostatic testing may be discontinued when the definition given in the relevant unit procedure for ending the activity is satisfied.
The acceptable P/T combinations are below and to the right of the limit curves. The limit curves include the limiting pressure differential between the point of system pressure measurement and the pressure on the reactor vessel region controlling the limit curve. However, the limit curves are not adjusted for possible instrument error and should not be used for operation (as identified in Note 1 on each applicable Figure).
SR 3.4.3.1 is modified by a Note that requires this SR to be performed only during system heatup operations with fuel in the reactor vessel. This SR refers to Figure 3.4.3-1 which provides applicable heatup limitations, including reactor coolant pump (RCP) operating restrictions and allowable heatup rates. Figure 3.4.3-1 Note 2 identifies that when the decay heat removal system is operating with no RCPs operating, the indicated DHR system return temperature to the reactor vessel is the appropriate temperature indicator.
SR 3.4.3.2 is modified by a Note that requires this SR to be performed only during system cooldown operations with fuel in the reactor vessel. This SR refers to Figure 3.4.3-2 which provides applicable cooldown limitations, including reactor coolant pump (RCP) operating restrictions and allowable cooldown rates. During system cooldown operations with fuel in the reactor vessel, the RCPs are eventually removed from service. Figure 3.4.3-2 Note 2 identifies that when the decay heat removal system is operating with no RCPs operating, the indicated decay heat removal system return temperature to the reactor vessel is the appropriate temperature indicator. Figure 3.4.3-2 Note 2 also indicates that a maximum step temperature change of 25 &deg;F is allowable when removing all RCPs from operation with the decay heat removal system operating. The step temperature change is defined as the reactor coolant temperature (prior to stopping all RCPs) minus the decay heat removal system return temperature to the reactor vessel (after stopping all RCPs). The step change of 25 &deg;F is applicable only during transition from RCP operation to DHR. This step change must be included when determining the cooldown rate.
SR 3.4.3.3 is modified by a Note that requires this SR to be performed only during system heatup and cooldown operations with no fuel in the reactor vessel. This SR refers to Figure 3.4.3-3. These curves are used during inservice hydrostatic testing that is performed in a defueled condition. The Notes on Figure 3.4.3-1 and Figure 3.4.3-2 are applicable to heatups and cooldowns performed within these limits.
ANO-1                                        B 3.4.3-7                          Amendment No. 215 Rev. 32,51,
 
RCS Loops - MODES 1 and 2 B 3.4.4 SURVEILLANCE REQUIREMENTS SR 3.4.4.1 This SR requires verification every 12 hours of the required number of loops in operation.
Verification includes flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal while maintaining the margin to DNB. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.The 12 hour interval has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions. In addition, control room indication and alarms will normally indicate loop status.
REFERENCES
: 1. SAR, Chapters 14 and 3A.
: 2. BAW-10103A, Revision 3, July 1977.
: 3. 10 CFR 50.36.
ANO-1                                        B 3.4.4-4                      Amendment No. 215 Rev.
 
RCS Loops - MODE 3 B 3.4.5 ACTIONS A.1 If one RCS loop is inoperable, redundancy for forced flow heat removal is lost. The Required Action is restoration of the RCS loop to OPERABLE status within a Completion Time of 72 hours. This time allowance is a justified period to be without the redundant non-operating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core.
B.1 If the Required Action and associated Completion Time of Condition A are not met, the unit must be brought to MODE 4. In MODE 4, the unit may be placed on the DHR System. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to achieve cooldown and depressurization from the existing unit conditions and without challenging unit systems.
C.1 and C.2 If no RCS loop is OPERABLE or a required RCS loop is not in operation, (no RCS loop is required to be in operation provided the conditions of the Note in the LCO section are met), all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be immediately suspended. Action to restore one RCS loop to operation shall be immediately initiated and continued until one RCS loop is restored to operation and to OPERABLE status. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for decay heat removal.
SURVEILLANCE REQUIREMENTS SR 3.4.5.1 This SR requires verification every 12 hours that the required loop (and pump) is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour interval has been shown by operating practice to be sufficient to regularly assess RCS loop status. In addition, control room indication and alarms will normally indicate loop status.
ANO-1                                        B 3.4.5-3                      Amendment No. 215 Rev.
 
RCS Loops - MODE 3 B 3.4.5 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.5.2 Verification that each required RCP is OPERABLE ensures that an RCS loop can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power availability to each required pump that is not in operation. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
REFERENCES
: 1. 10 CFR 50.36.
ANO-1                                        B 3.4.5-4                        Amendment No. 215 Rev.
 
RCS Loops - MODE 4 B 3.4.6 ACTIONS (continued)
B.1 and B.2 If no RCS or DHR loops are OPERABLE or a required loop is not in operation (no loop is required to be in operation provided the conditions of the Note in the LCO section are met), all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RCS or DHR loop to OPERABLE status and operation must be initiated. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must continue until one loop is restored to operation.
SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This Surveillance requires verification every 12 hours of the required DHR or RCS loop in operation to ensure forced flow is providing decay heat removal. Verification includes flow rate, temperature, or pump status monitoring. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12-hour interval has been shown by operating practice to be sufficient to regularly assess RCS loop status. In addition, control room indication and alarms will normally indicate loop status.
SR 3.4.6.2 Verification that each required pump is OPERABLE ensures that an RCS or DHR loop can be placed in operation if needed to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 7 days is considered reasonable in view of other administrative controls and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
REFERENCES
: 1. 10 CFR 50.36.
: 2. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
ANO-1                                        B 3.4.6-4                      Amendment No. 215 Rev. 50,
 
RCS Loops - MODE 5, Loops Filled B 3.4.7 SURVEILLANCE REQUIREMENTS SR 3.4.7.1 This SR requires verification every 12 hours that the required DHR loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour Frequency has been shown by operating practice to be sufficient to regularly assess degradation. In addition, control room indication and alarms will normally indicate loop status.
SR 3.4.7.2 Verifying the SGs are OPERABLE by ensuring their secondary side water levels are  20 inches ensures that redundant heat removal paths are available if the second DHR loop is not OPERABLE. If both DHR loops are OPERABLE, this Surveillance is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour Frequency has been shown by operating practice to be sufficient to regularly assess RCS loop status.
SR 3.4.7.3 Verification that each required DHR pump is OPERABLE ensures that a DHR loop can be placed in operation if needed to maintain decay heat removal and reactor coolant circulation. If the secondary side water level is  20 inches in both SGs, this Surveillance is not needed.
Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
REFERENCES
: 1. 10 CFR 50.36.
ANO-1                                        B 3.4.7-4                        Amendment No. 215 Rev. 30,
 
RCS Loops - MODE 5, Loops Not Filled B 3.4.8 ACTIONS (continued)
B.1, B.2, and B.3 If no required loop is OPERABLE or the required loop is not in operation, except as provided by Note 1 in the LCO, the Required Action requires immediate suspension of all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 or reduction of RCS water inventory and requires initiation of action to immediately restore one DHR loop to OPERABLE status and operation.
The Required Action for restoration does not apply to the condition of both loops not in operation when the exception Note in the LCO is in force. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operations for decay heat removal. The action to restore must continue until one loop is restored.
SURVEILLANCE REQUIREMENTS SR 3.4.8.1 This Surveillance requires verification every 12 hours that at least one loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour interval has been shown by operating practice to be sufficient to regularly assess RCS loop status.
SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that redundancy for heat removal is provided. The requirement also ensures that a DHR loop can be placed in operation if needed to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
REFERENCES
: 1. Generic Letter 88-17, October 17, 1988.
: 2. 10 CFR 50.36.
ANO-1                                        B 3.4.8-3                      Amendment No. 215 Rev. 30,
 
Pressurizer B 3.4.9 ACTIONS (continued)
B.1 and B.2 If the water level cannot be restored, reducing core power constrains heat input effects that drive pressurizer insurge that could result from an anticipated transient. By shutting down the reactor and reducing reactor coolant temperature to at least MODE 3, the potential thermal energy of the reactor coolant mass for mass and energy releases is reduced.
Six hours is a reasonable time based upon operating experience to reach MODE 3 from full power in an orderly manner and without challenging unit systems. Further pressure and temperature reduction to MODE 4 with RCS temperature  259 &deg;F places the unit into a MODE where the LCO is not applicable. The 24 hour Completion Time to reach the non-applicable MODE is reasonable based upon operating experience.
C.1 If the required pressurizer heaters are inoperable, restoration is required in 72 hours. The Completion Time of 72 hours is reasonable considering the anticipation that a demand caused by loss of offsite power will not occur in this period. Pressure control may be maintained during this time using non-ES bus powered heaters.
D.1 and D.2 If the Required Action and associated Completion Time are not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 3 within 6 hours and to MODE 4 within the following 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. Similarly, the Completion Time of 12 hours to reach MODE 4 is reasonable based on operating experience to achieve power reduction from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE REQUIREMENTS SR 3.4.9.1 This SR requires that pressurizer water level be maintained below the upper limit to provide a minimum space for a steam bubble. The value specified for pressurizer level does not contain an allowance for instrument error. Therefore, additional allowances for instrument uncertainties must be provided in the implementing procedures. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12 hour interval has been shown by operating practice to be sufficient to regularly assess the level for any deviation and verify that operation is within safety analyses assumptions. Alarms are also available for early detection of abnormal level.
ANO-1                                          B 3.4.9-4                      Amendment No. 215 Rev. 35,51,
 
Pressurizer B 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.9.2 The SR requires sufficient pressurizer heaters which are connected to an ES bus verified to be capable of providing the required capacity (this may be performed by testing the power supply output and by performing an electrical check on heater element continuity and resistance). The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months is considered adequate to detect heater degradation and has been shown by operating experience to be acceptable.
REFERENCES
: 1. NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," July 1979.
: 2. 10 CFR 50.36, Technical Specifications.
ANO-1                                      B 3.4.9-5                      Amendment No. 215 Rev. 35,51,
 
LTOP System B 3.4.11 ACTIONS (continued)
C.1 and D.1 (continued)
These Completion Times also consider these activities can be accomplished in these time periods. A limiting LTOP event is not likely in those times.
Some ERV testing or maintenance can only be performed at unit shutdown. Such activity is permitted if Required Action D.1 is taken to compensate for required ERV unavailability.
E.1 With the LTOP requirements not met for any reason other than cited in Condition A through D, action must be initiated to restore compliance immediately. The immediate Completion Time reflects the urgency of quickly proceeding with the Required Actions.
SURVEILLANCE REQUIREMENTS SR 3.4.11.1 Verification of the pressurizer level at  180 inches by observing control room or other indications ensures that the unit is not in a water solid condition and that a cushion of sufficient size is available to reduce the rate of pressure increase from potential transients (Ref. 11). This parameter does not contain allowances for instrument error.
The 30-minute Surveillance Frequency during heatup and cooldown must be performed for the LCO Applicability period when temperature changes can cause pressurizer level variations.
This Frequency may be discontinued when these evolutions are complete, as defined in unit procedures. Thereafter, the Surveillance is required at 12-hour intervals.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThese Frequencies are shown by operating practice sufficient to regularly assess indications of potential degradation and verify operation within the safety analysis.
SR 3.4.11.2 and SR 3.4.11.3 Verifications must be performed that the HPI is deactivated, and each pressurized CFT is isolated. These Surveillances ensure the minimum coolant input capability will not create an RCS overpressure condition to challenge the LTOP. The Surveillances are required at 12-hour intervals.
The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe 12-hour intervals are shown by operating practice to be sufficient to assess coolant input capability and verify operation within the safety analysis.
ANO-1                                        B 3.4.11-6                        Amendment No. 215 Rev. 51,
 
LTOP System B 3.4.11 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.11.4 OPERABLE pressure relief capability must be provided to prevent overpressurization due to inadvertent full makeup system operation. Such a vent keeps the pressure from full makeup flow within the LCO limit. OPERABLE pressure relief capability may be provided by an OPERABLE ERV, or by depressurizing the RCS and providing an alternate RCS vent path.
For the ERV to be considered OPERABLE, its block valve must be open, its lift setpoint must be set at  508 psig, testing must have proven its ability to open at that setpoint, and motive power must be available to the two valves and their control circuits. The parameter value of 508 psig does not contain allowances for instrument uncertainty. Additional allowances for instrument uncertainty are contained in the implementing procedures.
With the RCS depressurized, acceptable alternate vent paths include: a) removing a pressurizer safety valve; b) locking the ERV in the open position and disabling its block valve in the open position; c) removing a SG primary manway; c) removing a SG primary hand hole cover; d) removing all control rod drive top closure assemblies (excluding reactor vessel level probe); and e) removing a pressurizer manway.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramFor a vent path not locked open, the Frequency is every 12 hours. For a locked open vent path, the required Frequency is every 31 days.
The Frequency intervals are considered adequate based on operating practice to determine adequacy of pressure relief capability and verify operation within the safety analysis.
SR 3.4.11.5 The performance of a CHANNEL CALIBRATION is required every 18 months. The CHANNEL CALIBRATION for the LTOP ERV opening logic, including the ERV setpoint, ensures that the ERV will be actuated at the appropriate RCS pressure by verifying the accuracy of the instrument string. The calibration can only be performed in shutdown.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency considers a typical refueling cycle and industry accepted practice.
ANO-1                                        B 3.4.11-7                        Amendment No. 215 Rev. 51,
 
RCS Specific Activity B 3.4.12 SURVEILLANCE REQUIREMENTS SR 3.4.12.1 SR 3.4.12.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.
Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The 7 day Frequency considers the low probability of a gross fuel failure during this time. The Surveillance is modified by a note requiring performance in MODES 1 and 2, and in MODE 3 with RCS average temperature 500&deg;F. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7 day Frequency is based on the low probability of a gross fuel failure during that time period.
Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and I-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.12.1 calculation. If a specific noble gas nuclide listed in the definition of DEX is not detected, it should be assumed to be present at the minimum detectable activity.
SR 3.4.12.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 14 day Frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days.
REFERENCES
: 1. 10 CFR 50.67.
: 2. ANO-1 Operating License Amendment 238 (1CNA100901), dated October 21, 2009.
ANO-1                                          B 3.4.12-4                      Amendment No. 215 Rev. Rev. 22,33,38,
 
RCS Operational LEAKAGE B 3.4.13 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.13.1 (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 72 hour Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.
SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.16, Steam Generator Tube Integrity, should be evaluated.
The 150 gallons per day limit is measured at room temperature as described in Reference 8.
The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance Frequency of 72 hours is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with EPRI Guidelines (Ref. 8).
REFERENCES
: 1. SAR, Section 1.4, GDC 30.
: 2. Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973.
: 3. Information Submittal - Comparison of ANO-1 RCS Leak Detection Systems to Regulatory Guide 1.45 (1CAN108607), dated October 14, 1986.
: 4. SAR, Chapter 14.
: 5. SAR, Section 4.2.3.8.
: 6. 10 CFR 50.36.
: 7. NEI 97-06, Steam Generator Program Guidelines.
ANO-1                                      B 3.4.13-5                        Amendment No. 215 Rev. 7,
 
RCS PIV Leakage B 3.4.14 SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.14.2, SR 3.4.14.3, SR 3.4.14.4, and SR 3.4.14.5 Verifying that the DHR autoclosure interlocks are OPERABLE ensures that RCS pressure will not over pressurize the DHR system. The interlock(s) that prevent the valves from being opened and that close the valves are designed to protect the DHR System from gross overpressurization. Although the specified values include certain process measurement uncertainties, additional allowances for instrument uncertainty are contained in the implementing procedures.
The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and on the potential for an unplanned transient if the Surveillance was performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.
REFERENCES
: 1.    "Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves," issued April 20, 1981.
: 2. NUREG-75/014, Reactor Safety Study, Appendix V, October 1975.
: 3. NUREG-0677, The Probability of Intersystem LOCA: Impact Due to Leak Testing and Operational Changes, May 1980.
: 4. 10 CFR 50.36.
: 5. ASME OM Code 2004 Edition through 2006 Addenda and Code Case OMN-20 (Inservice Test Frequency).
: 6. 10 CFR 50.55a(f).
ANO-1                                        B 3.4.14-5                      Amendment No. 215 Rev. 37,58,60,
 
RCS Leakage Detection Instrumentation B 3.4.15 ACTIONS (continued)
D.1 and D.2 (continued)
Required Action D.2 is modified by a second Note. Note 2 states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate.
LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
E.1 With both required monitors inoperable, no indicated means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.
SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required reactor building atmosphere radioactivity monitor. The check gives reasonable confidence that each channel is operating properly. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions.
SR 3.4.15.2 SR 3.4.15.2 requires the performance of a CHANNEL FUNCTIONAL TEST of the required reactor building atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm function and relative accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 92 days considers instrument reliability, and operating experience has shown it proper for detecting degradation.
SR 3.4.15.3 and SR 3.4.15.4 These SRs require the performance of a CHANNEL CALIBRATION for each of the required RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside the reactor building. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months is a typical refueling cycle and considers channel reliability. Additionally, operating experience has shown this Frequency is acceptable.
ANO-1                                        B 3.4.15-5                      Amendment No. 215 Rev. 22,24,44,50,
 
CFTs B 3.5.1 SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Verification every 12 hours that each CFT isolation valve is fully open ensures that the CFTs are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in accident analysis assumptions not being met. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramA 12 hour Frequency is considered reasonable in view of administrative controls that ensure that a mispositioned isolation valve is unlikely.
SR 3.5.1.2 and SR 3.5.1.3 Verification every 12 hours of each CFT's nitrogen cover pressure and the borated water volume is sufficient to ensure adequate injection during a LOCA. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramDue to the static nature of these parameters, a 12 hour Frequency usually allows the operator to identify changes before the limits are reached. Operating experience has shown that this Frequency is appropriate for early detection and correction of off normal trends.
SR 3.5.1.4 This Surveillance once every 31 days is reasonable to verifiesy that the CFT boron concentration is within the required limits, because the static nature of this parameter limits the ways in which the concentration can be changed. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage.
Sampling of the affected CFT within 12 hours after a 0.2 ft volume increase will identify whether inleakage from the RCS has caused a reduction in boron concentration to below the required limit. The 0.2 ft increase represents approximately 102 gallons increase in volume. It is not necessary to verify boron concentration if the added water inventory is from a borated water source of known concentration  2270 ppm, such as the borated water storage tank (BWST),
because the water is within CFT boron concentration requirements. Similarly, it would not be necessary to sample the CFT following inventory additions from the CFT makeup tank if sampling has determined that the added inventory had a boron concentration within the CFT requirements. This is consistent with the recommendations of NUREG-1366 (Ref. 4).
SR 3.5.1.5 Removing power from each CFT isolation valve operator ensures that an active failure could not result in the undetected closure of a CFT motor operated isolation valve coincident with a LOCA. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramSince power is removed under administrative control, the 31 day Frequency will provide adequate assurance that the power is removed.
ANO-1                                          B 3.5.1-5                        Amendment No. 215 Rev. 51,
 
ECCS - Operating B 3.5.2 SURVEILLANCE REQUIREMENTS For inservice testing periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the inservice testing periods by up to 25%. For inservice testing periods of greater than 2 years, OMN-20 allows an extension of up to 6 months.
SR 3.5.2.1 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is appropriate because the valves are operated under administrative control, and an inoperable valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.
SR 3.5.2.2 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME OM Code (Ref. 7). This testing confirms component OPERABILITY, trends performance, and detects incipient failures by indicating abnormal performance. SRs are specified in the INSERVICE TESTING PROGRAM, which encompasses the ASME OM Code.
SR 3.5.2.3 This SR demonstrates that each automatic ECCS valve actuates to the required position on an actual or simulated ESAS signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the ESAS testing, and equipment performance is monitored as part of the INSERVICE TESTING PROGRAM.
SR 3.5.2.4 The intent of this SR is to verify that the ECCS pumps are capable of automatically starting on an ESAS signal. Because of the system design configuration and the limitations imposed on pump operation during the unit conditions when this test would be conducted, this verification must be conducted through a series of sequential, overlapping or total steps in order to demonstrate functionality.
ANO-1                                          B 3.5.2-6                        Amendment No. 215 Rev. 30,37,58,
 
ECCS - Operating B 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.4 (continued)
SR 3.5.2.4 demonstrates that each ECCS pump would be capable of starting by verifying that its breaker closes on receipt of an actual or simulated ESAS signal. SR 3.5.2.4 works in conjunction with the INSERVICE TESTING PROGRAM (SR 3.5.2.2) which periodically verifies the ability of the pumps to start and operate within limits, and the ESAS actuation logic testing which periodically verifies the ability of the ESAS to sense, process and generate an actuation signal.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.
SR 3.5.2.5 Periodic inspections of the reactor building sump suction inlet ensure that it is unrestricted and stays in proper operating condition. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on the need to perform this Surveillance during a unit outage. Operating experience has shown this Frequency to be acceptable to detect abnormal degradation.
REFERENCES
: 1. SAR, Section 6.
: 2. Letter from A. C. Thadani (NRC) to P. S. Walsh (BWOG) dated March 9, 1993.
: 3. 10 CFR 50.46.
: 4. SAR, Section 14.2.2.5.2.
: 5. 10 CFR 50.36.
: 6. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.
: 7. ASME OM Code 2004 Edition through 2006 Addenda and Code Case OMN-20 (Inservice Test Frequency).
: 8. Condition Report CR-ANO-1-2009-0997.
ANO-1                                          B 3.5.2-7                      Amendment No. 215 Rev. 30,37,58,60,
 
BWST B 3.5.4 SURVEILLANCE REQUIREMENTS SR 3.5.4.1 Verification every 24 hours that the BWST water temperature is within the specified temperature band ensures that the fluid will not freeze and that the fluid temperature will not be hotter than assumed in the safety analysis. These parameter values are considered to be nominal values and do not contain an allowance for instrument uncertainty. No additional allowances for instrument uncertainty are required to be included in the implementing procedures. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 24-hour Frequency is sufficient to identify a temperature change that would approach either temperature limit.
The SR is modified by a Note that requires the Surveillance to be performed only when ambient air temperatures are outside the operating temperature limits of the BWST. With ambient temperatures within this band, the BWST temperature should not exceed the limits.
SR 3.5.4.2 Verification every 7 days that the BWST level is  38.4 feet and  42 feet ensures that a sufficient initial supply is available for injection and to support continued ECCS pump operation on recirculation. These levels correspond to volumes of approximately 375,096 gallons and 405,090 gallons, respectively. These parameter values do not contain an allowance for instrument uncertainty. Additional allowances for instrument uncertainty are included in the implementing procedures. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramSince the BWST level is normally stable, a 7-day Frequency has been shown to be appropriate through operating experience.
SR 3.5.4.3 Verification every 7 days that the boron concentration of the BWST fluid is  2270 ppm and 2670 ppm ensures that the reactor will remain adequately shutdown following a LOCA. These parameter values do not contain an allowance for instrument uncertainty. Additional allowances for instrument uncertainty are included in the implementing procedures. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramSince the BWST level is normally stable, a 7-day sampling Frequency is appropriate and has been shown to be acceptable through operating experience.
REFERENCES
: 1. SAR, Section 6.1.
: 2. Letter from A. C. Thadani (NRC) to P. S. Walsh (BWOG) dated March 9, 1993.
: 3. 10 CFR 50.36.
ANO-1                                            B 3.5.4-5                    Amendment No. 215 Rev. 1,50,
 
Reactor Building Air Locks B 3.6.2 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident reactor building pressure, closure of either door will support the reactor building OPERABILITY. Thus, the door interlock feature supports the reactor building OPERABILITY while the air lock is being used for personnel transit in and out of the reactor building. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramDue to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the reactor building air lock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 18 months. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage, and the potential for loss of reactor building OPERABILITY if the Surveillance were performed with the reactor at power.
The 18 month Frequency for the interlock is justified based on generic operating experience.
The 18 month Frequency is based on engineering judgment and is considered adequate given that the interlock is not expected to be challenged during use of the airlock.
REFERENCES
: 1. 10 CFR 50, Appendix J, Option B.
: 2. SAR, Chapter 14.
: 3. SAR, Chapter 5.
: 4. 10 CFR 50.36.
: 5. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
ANO-1                                          B 3.6.2-6                      Amendment No. 215 Rev. 50,
 
Reactor Building Isolation Valves B 3.6.3 ACTIONS (continued)
D.1 and D.2 Remaining within the Applicability of the LCO is acceptable because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 7). The release of stored energy to the Reactor Building in the event of an accident in MODE 4 is substantially less than the energy release assumed due to an accident at power. Therefore, the challenge to containment isolation valves is substantially reduced. Because of the reduction in RCS pressure and temperature in MODE 4, the likelihood of an event is also reduced. In addition, there are more accident mitigation systems available and there is more redundancy and diversity in core heat removal mechanisms in MODE 4 than in MODE 5. However, voluntary entry into MODE 5 may be made as it is also an acceptable low-risk state.
Required Action D.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
SURVEILLANCE REQUIREMENTS SR 3.6.3.1 Each 24-inch reactor building purge isolation valve in the purge system supply and exhaust is required to be periodically verified closed at 31 day intervals. This Surveillance is designed to ensure that a gross breach of the reactor building is not caused by an inadvertent opening of a reactor building purge valve. Detailed analysis of the purge valves failed to conclusively demonstrate their ability to close during a LOCA in time to limit offsite doses. Therefore, these valves are required to be in the closed position during MODES 1, 2, 3, and 4. A reactor building purge valve that is closed must have motive power to the valve operator removed. This can be accomplished by removing the valve handswitch key. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with other reactor building isolation valves discussed in SR 3.6.3.2.
SR 3.6.3.2 This SR requires verification that each reactor building isolation manual valve and blind flange located outside the reactor building and not locked, sealed, or otherwise secured, and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the reactor building boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that those reactor building isolation valves outside the reactor building and capable of being mispositioned are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramSince verification of valve position for the reactor building isolation valves outside the reactor building is relatively easy, the 31-day Frequency was chosen to provide added assurance of the correct positions.
ANO-1                                          B 3.6.3-7                        Amendment No. 215 Rev. 28,39,50,
 
Reactor Building Isolation Valves B 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.5 Automatic reactor building isolation valves close on a reactor building isolation signal to prevent leakage of radioactive material from the reactor building following a DBA. This SR ensures that each automatic reactor building isolation valve will actuate to its isolation position on a reactor building isolation signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 18-month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES
: 1. SAR, Chapter 5.
: 2. SAR, Chapter 14.
: 3. 10 CFR 50.36.
: 4. SAR, Table 5-1.
: 5. Generic Letter 91-08, Removal of Component Lists from Technical Specifications.
: 6. Condition Report CR-ANO-1-2010-2515.
: 7. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
: 8. ASME OM Code 2004 Edition through 2006 Addenda and Code Case OMN-20 (Inservice Test Frequency).
ANO-1                                        B 3.6.3-9                        Amendment No. 215 Rev. 28,39,50,60,
 
Reactor Building Pressure B 3.6.4 ACTIONS (continued)
B.1 and B.2 (continued)
Remaining within the Applicability of the LCO is acceptable because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 5). The release of stored energy to the Reactor Building in the event of an accident in MODE 4 is substantially less than the energy release assumed due to an accident at power. Therefore, the challenge to the containment systems due to an increase in containment pressure is substantially reduced. Because of the reduction in RCS pressure and temperature in MODE 4, the likelihood of an event is also reduced. In addition, there are more accident mitigation systems available and there is more redundancy and diversity in core heat removal mechanisms in MODE 4 than in MODE 5. However, voluntary entry into MODE 5 may be made as it is also an acceptable low-risk state.
Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
SURVEILLANCE REQUIREMENTS SR 3.6.4.1 Verifying that the reactor building pressure is within limits ensures that operation remains within the limits assumed in the ECCS and the reactor building analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12-hour Frequency of this SR was developed after taking into consideration operating experience related to trending of the reactor building pressure variations during the applicable MODES. Furthermore, the 12-hour Frequency is considered adequate in view of other indications available in the control room to alert the operator to an abnormal reactor building pressure condition.
REFERENCES
: 1. SAR, Chapter 14.
: 2. SAR, Chapter 5.
: 3. 10 CFR 50, Appendix K.
: 4. 10 CFR 50.36.
: 5. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
ANO-1                                          B 3.6.4-3                      Amendment No. 215 Rev. 50,
 
Reactor Building Spray and Cooling Systems B 3.6.5 ACTIONS (continued)
G.1 With two reactor building spray trains inoperable in MODE 1 or 2, or any combination of three or more reactor building spray and reactor building cooling trains inoperable in MODE 1 or 2, or one required reactor building spray train and one required reactor building cooling train inoperable in MODE 3 or 4, then LCO 3.0.3 must be entered immediately. The first part of this Condition addresses the loss of Spray Additive System support which would result from two inoperable reactor building spray trains in MODE 1 or 2. The second part of this Condition considers the loss of adequate reactor building cooling capacity in MODE 1 or 2 which would result from the loss of three or more of the four RB spray and RB cooling trains. Finally, the third part of this Condition addresses loss of reactor building cooling capability in MODES 3 and 4 when only one train of RB spray and one train of RB cooling are required.
SURVEILLANCE REQUIREMENTS SR 3.6.5.1 Verifying the correct alignment for manual, power operated, and automatic valves in the reactor building spray flow path provides assurance that the proper flow paths will exist for the Reactor Building Spray System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification, through a system walkdown or control room indication, that those valves outside the reactor building and capable of potentially being mispositioned are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program SR 3.6.5.2 Operating each required reactor building cooling train fan unit for  15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. This SR is performed by starting (unless operating) each operational cooling fan from the control room.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency was developed considering the known reliability of the fan units and controls, the redundancy available, and the low probability of a significant degradation of the reactor building cooling trains occurring between surveillances and has been shown to be acceptable through operating experience.
SR 3.6.5.3 Verifying that a service water flow rate of 1200 gpm is provided to each required reactor building cooling train provides assurance that the original design flow rate is being achieved and that the service water flow rate is not degrading (Ref. 3). Assurance that the flow doesnt degrade by biological fouling between surveillances is provided by the addition of a biocide to the Service Water System whenever the service water temperature is between 60 &deg;F and 80 &deg;F. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency was developed considering the known reliability of the system, the redundancy available, and the low probability of a significant degradation of flow occurring between surveillances.
ANO-1                                          B 3.6.5-6                      Amendment No. 215 Rev. 30,37,58,
 
Reactor Building Spray and Cooling Systems B 3.6.5 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.5.4 Verifying that each required reactor building spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Acceptable performance will be indicated if the pump starts, operates for fifteen minutes, and the discharge pressure and flow rate are within +/- 10 % of a point on the pump head curve. Flow and differential pressure are measured during normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 5). Since the Reactor Building Spray System pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms the discharge pressure and flow rate are within
+/- 10 % of a point on the pump head curve and is indicative of overall pump performance. Such inservice tests confirm component OPERABILITY, trend performance, and may detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the INSERVICE TESTING PROGRAM.
For inservice testing periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the inservice testing periods by up to 25%. For inservice testing periods of greater than 2 years, OMN-20 allows an extension of up to 6 months.
SR 3.6.5.5 and SR 3.6.5.6 These SRs require verification that each automatic reactor building spray valve actuates to its correct position and that each reactor building spray pump starts upon receipt of an actual or simulated actuation signal. The SRs are considered satisfactory if visual observation and control board indication verifies that all components have responded to the actuation signal properly. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls. During testing of the spray pump, the reactor building isolation valve in the spray line is closed with its breaker open to prevent spraying the reactor building. After spray pump performance is verified, the pump is stopped. Its breaker is racked down to prevent restart. Power is then restored to the reactor building isolation valve for valve testing. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillances when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.6.5.7 This SR requires verification by control board indication that each required reactor building cooling train actuates upon receipt of an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency has been shown to be acceptable through operating experience. See SR 3.6.5.5 and SR 3.6.5.6, above, for further discussion of the basis for the 18 month Frequency.
ANO-1                                          B 3.6.5-7                        Amendment No. 215 Rev. 30,37,58,
 
Spray Additive System B 3.6.6 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6.1 (continued)
This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown or control room indication, that those valves outside the reactor building capable of potentially being mispositioned are in the correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program SR 3.6.6.2 To provide the most effective iodine removal, the reactor building spray should be an alkaline solution. Since the BWST contents are normally acidic, the NaOH tank must provide a sufficient volume of NaOH to adjust pH for all water injected. This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System. The NaOH tank solution minimum volume of 9000 gallons corresponds to a tank level of approximately 26 feet at a temperature of 77F and a NaOH concentration of 6.0 wt%. This parameter does not contain an allowance for instrument uncertainty. Additional allowances for instrument uncertainty are contained in the implementing procedures. The minimum NaOH tank volume preserves the required NaOH solution contribution from the tank to the post-LOCA minimum sump level. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 184 day Frequency is based on the low probability of an undetected change in tank volume occurring during the SR interval (the tank is isolated during normal unit operations). Tank level is also indicated and alarmed in the control room, such that there is a high confidence that a substantial change in level would be detected.
SR 3.6.6.3 This SR provides verification of the NaOH concentration in the NaOH tank and is sufficient to ensure that the spray solution being injected into the reactor building is at the correct pH level.
The concentration of NaOH in the NaOH tank must be determined by chemical analysis. This parameter is considered to be a nominal value; therefore, additional allowance for instrument uncertainty need not be applied. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 184 day Frequency is sufficient to ensure that the concentration of NaOH in the tank remains within the established limits. This is based on the low likelihood of an uncontrolled change in concentration (the tank is normally isolated) and the probability that any substantial variance in tank volume will be detected.
SR 3.6.6.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
ANO-1                                          B 3.6.6-4                        Amendment No. 215 Rev. 26,
 
MSIVs B 3.7.2 SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.2.1 (continued)
This test is normally conducted in MODE 3, with the unit at operating temperature and pressure.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows delaying testing until MODE 3 in order to establish conditions consistent with those under which the acceptance criterion was generated.
For inservice testing periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the inservice testing periods by up to 25%. For inservice testing periods of greater than 2 years, OMN-20 allows an extension of up to 6 months.
SR 3.7.2.2 This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of MSIV testing is every 18 months. The 18 month Frequency for testing is based on the refueling cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.
This SR is modified by two Notes. The first Note allows entry into and operation in MODE 3 prior to performing the SR. This allows delaying testing until MODE 3 in order to establish conditions consistent with those under which the acceptance criterion was established.
SR 3.7.2.2 is also modified by a second Note which indicates that the automatic closure capability is not required to be met when SG pressure is < 750 psig. At < 750 psig, the main steam line isolation Function of EFIC may be disabled to prevent automatic actuation on low steam generator pressure during a unit shutdown.
REFERENCES
: 1. SAR, Section 14.2.
: 2. SAR, Section 7.1.4.
: 3. 10 CFR 50.36.
: 4. ASME OM Code 2004 Edition through 2006 Addenda and Code Case OMN-20 (Inservice Test Frequency).
ANO-1                                        B 3.7.2-4                      Amendment No. 215 Rev. 58,60,
 
MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves B 3.7.3 SURVEILLANCE REQUIREMENTS SR 3.7.3.1 This SR verifies that the closure time of each MFIV, Main Feedwater Block Valve, Low Load Feedwater Control Valve and Startup Feedwater Control Valve is as specified in the INSERVICE TESTING PROGRAM.
The MFIV, Main Feedwater Block Valve, Low Load Feedwater Control Valve and Startup Feedwater Control Valve isolation time is assumed in the accident and reactor building analyses. This Surveillance is normally performed prior to returning the unit to power operation, e.g., during MODE 3, following a refueling outage. The MFIVs, Main Feedwater Block Valves, Low Load Feedwater Control Valves and Startup Feedwater Control Valves are not tested at power since even a part stroke exercise increases the risk of a valve closure with the unit generating power.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The Frequency for this SR is in accordance with the INSERVICE TESTING PROGRAM.
For inservice testing periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the inservice testing periods by up to 25%. For inservice testing periods of greater than 2 years, OMN-20 allows an extension of up to 6 months.
SR 3.7.3.2 This SR verifies that each MFIV, Main Feedwater Block Valve, Low Load Feedwater Control Valve and Startup Feedwater Control Valve can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program The Frequency for this SR is every 18 months. The 18 month Frequency for testing is based on the refueling cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.
This SR is modified by two Notes. The first Note allows entry into and operation in MODE 3 prior to performing the SR. This allows delaying testing until MODE 3 in order to establish conditions consistent with those under which the acceptance criterion was established.
SR 3.7.3.2 is also modified by a second Note which indicates that the automatic closure capability is not required to be met when the SG pressure is < 750 psig. At < 750 psig, the main steam line isolation Function of EFIC may be disabled to prevent automatic actuation on low SG pressure during a unit shutdown.
ANO-1                                        B 3.7.3-5                      Amendment No. 215 Rev. 40,58,
 
Secondary Specific Activity B 3.7.4 LCO (continued)
Monitoring the specific activity of the secondary coolant ensures that, when secondary specific activity limits are exceeded, appropriate actions are taken, in a timely manner, to place the unit in an operational MODE that would minimize the radiological consequences of a DBA.
APPLICABILITY In MODES 1, 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam releases to the atmosphere.
In MODES 5 and 6, the steam generators are not being used for heat removal. Both the RCS and steam generators are at low pressure and primary to secondary LEAKAGE is minimal.
Therefore, secondary specific activity is not a concern.
ACTIONS A.1 and A.2 DOSE EQUIVALENT I-131 exceeding the allowable value in the secondary coolant contributes to increased post accident doses. If secondary specific activity cannot be restored to within limits within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 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 REQUIREMENTS SR 3.7.4.1 This SR verifies that the secondary specific activity is within the limits of the accident analysis assumptions. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT I-131, confirms the analysis assumptions are met. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is based on the detection of increasing trends of the level of DOSE EQUIVALENT I-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.
REFERENCES
: 1. 10 CFR 50.67.
: 2. Safety Evaluation Report for ANO-1 License Amendment No. 2, 1CNA057502, dated May 9, 1975.
ANO-1                                        B 3.7.4-2                          Amendment No. 215 Rev. 33,
 
EFW System B 3.7.5 ACTIONS (continued)
E.1 In MODE 4, either the steam generator loops or the DHR loops can be used to provide heat removal, which is addressed in LCO 3.4.6, RCS Loops - MODE 4. With the required EFW train inoperable, action must be taken to immediately restore the inoperable train to OPERABLE status.
SURVEILLANCE REQUIREMENTS SR 3.7.5.1 Verifying the correct alignment for manual, power operated, and automatic valves in the EFW water and steam supply flow paths provides assurance that the proper flow paths exist for EFW operation. Correct alignment for automatic valves may be other than the post-accident position provided the valve is otherwise OPERABLE. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since those valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is based on the procedural controls governing valve operation, and ensures correct valve positions.
SR 3.7.5.2 Verifying that each EFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that EFW pump performance has not degraded below the established acceptance criteria during the cycle. Flow and differential head are indicators of pump performance required by the ASME OM Code (Ref. 5). Because it is undesirable to introduce cold EFW into the steam generators while they are operating, this test may be performed on a test flow path.
This test is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. Performance of inservice testing as discussed in the ASME OM Code (Ref. 5) and the INSERVICE TESTING PROGRAM satisfies this requirement.
This SR is modified by a Note indicating that the SR may be deferred until suitable test conditions are established. This deferral is required because there may be insufficient steam pressure to perform the test.
For inservice testing periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the inservice testing periods by up to 25%. For inservice testing periods of greater than 2 years, OMN-20 allows an extension of up to 6 months.
ANO-1                                          B 3.7.5-5                        Amendment No. 215 Rev. 22,37,52,58,
 
EFW System B 3.7.5 SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.5.3 This SR verifies that EFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an EFIC system signal by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 18-month Frequency is also acceptable based on operating experience and design reliability of the equipment.
This SR is modified by a Note which states that the SR is not required to be met in MODE 4 when the steam generator is being relied upon for heat removal. In MODE 4, the heat removal requirements would be less providing more time for operator action to manually start the required EFW pump.
SR 3.7.5.4 This SR verifies that each EFW pump starts in the event of any accident or transient that generates an EFIC signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
This SR is modified by a Note which states that the SR is not required to be met in MODE 4 when the steam generator is being relied upon for heat removal. In MODE 4, the heat removal requirements would be less providing more time for operator action to manually start the required EFW pump.
SR 3.7.5.5 This SR ensures that the EFW system is properly aligned by verifying the position of manual valves in the flow paths to each steam generator prior to entering MODE 2 after more than 30 days in any combination of MODE 5 or 6 or defueled. OPERABILITY of EFW flow paths must be demonstrated before sufficient core heat is generated that would require the operation of the EFW System during a subsequent shutdown. The Frequency is reasonable in view of other administrative controls, such as SR 3.7.5.1, to ensure that the flow paths are OPERABLE.
To further ensure EFW System alignment, flow path OPERABILITY is verified, following extended outages to determine no misalignment of manual valves has occurred. This SR ensures that the flow path from the QCST to the steam generator is properly aligned.
ANO-1                                        B 3.7.5-6                        Amendment No. 215 Rev. 22,52,
 
EFW System B 3.7.5 SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.5.6 This SR ensures that the EFW flow path to each steam generator is open and that water reaches the steam generators from the EFW system. This test is performed during shutdown to minimize thermal cycles to the emergency feedwater nozzles on the steam generator due to the lower temperature of the emergency feedwater. The motor-driven EFW pump is specified because of its availability at the low steam generator pressure conditions that exist in the shutdown condition. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
REFERENCES
: 1. SAR, Section 7.1.4.
: 2. SAR, Section 10.4.8.
: 3. NRC Letter dated January 12, 1981, (1CNA018103).
: 4. 10 CFR 50.36.
: 5. ASME OM Code 2004 Edition through 2006 Addenda and Code Case OMN-20 (Inservice Test Frequency).
ANO-1                                        B 3.7.5-7                      Amendment No. 215 Rev. 22,37,52,58,60,
 
QCST B 3.7.6 APPLICABILITY In MODES 1, 2, 3, and 4 when a steam generator is being relied upon for heat removal, the QCST is required to be OPERABLE.
In MODE 4 when a steam generator is not being relied upon for heat removal, and in MODES 5 and 6, the QCST is not required because the EFW System is not required.
ACTIONS A.1 and A.2 As an alternative to unit shutdown, the OPERABILITY of the backup water supply (SWS) should be verified within 4 hours and once every 12 hours thereafter. The OPERABILITY of the SWS backup feedwater supply must include verification, by administrative means, of the OPERABILITY of the SWS flow path to the EFW pumps. The QCST must be restored to OPERABLE status within 7 days because the backup supply may be performing this function in addition to its normal functions. The 4-hour Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the backup water supply. Additionally, verifying the backup water supply every 12 hours is adequate to ensure the backup water supply continues to be available. The 7-day Completion Time is reasonable, based on an OPERABLE SWS backup water supply being available, and the low probability of an event occurring during this time period, requiring the use of water from the QCST.
B.1 and B.2 If the Required Action and associated Completion Times are not met, the unit must be placed in a MODE in which the LCO does not apply, with the DHR System in operation. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance on steam generators for heat removal, within 24 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 REQUIREMENTS SR 3.7.6.1 This SR verifies that the QCST contains the required volume of cooling water. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12-hour Frequency is based on operating experience and the need for operator awareness of unit evolutions that may affect the QCST inventory between checks. The 12-hour Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to abnormal deviations in QCST levels.
ANO-1                                        B 3.7.6-2                      Amendment No. 215 Rev. 45,
 
SWS B 3.7.7 SURVEILLANCE REQUIREMENTS SR 3.7.7.1 Verifying the correct alignment for manual, power operated, and automatic valves in the SWS flow path provides assurance that the proper flow paths exist for SWS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR 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 also does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31-day Frequency is based on the existence of procedural controls governing valve operation, and ensures correct valve positions.
This SR is modified by a Note indicating that the isolation of components or systems supported by the SWS does not affect the OPERABILITY of the SWS. However, such isolation may render those components inoperable.
SR 3.7.7.2 The SR verifies proper automatic operation of the SWS valves. The SWS is a normally operating system that cannot be fully actuated as part of the normal testing. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and on the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18-month Frequency.
Therefore, the Frequency is acceptable from a reliability standpoint.
SR 3.7.7.3 This SR requires verification that the normally operating SWS pumps (A and C) automatically restart following restoration of power to the respective bus. In addition, the B SWS pump, normally in the standby condition, must be verified to start to support each SWS train for which it is expected to be aligned upon associated ES actuation (with time delay) with simulated failure of the normally operating pump for that train.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at an 18-month Frequency.
Therefore, the Frequency is acceptable from a reliability standpoint.
ANO-1                                          B 3.7.7-4                        Amendment No. 215 Rev. 41,46,50,
 
ECP B 3.7.8 ACTIONS (continued)
B.1 and B.2 If the ECP is inoperable, the unit must be placed in a MODE in which the LCO does not apply.
To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 5 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 REQUIREMENTS SR 3.7.8.1 This SR (together with SR 3.7.8.3 and SR 3.7.8.4) verifies that adequate long term (30 days) cooling inventory is available. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 24-hour Frequency is based on operating experience related to the trending of the ECP level during the applicable MODES. This SR verifies that the ECP indicated water level is  5.2 ft, which is sufficient to ensure a water volume of 70 acre-feet when crediting operator action to initiate makeup to the ECP upon a loss of Dardanelle Reservoir event (described in the Applicable Safety Analyses section above). The 5.2-foot minimum level requirement includes measurement, calculation, and other uncertainties.
SR 3.7.8.2 This SR provides assurance that the heat sink for the SWS can dissipate the maximum accident or normal heat loads for 30 days following the design basis event. The temperature, measured at the point of discharge from the ECP, is considered a conservative average of total ECP conditions since solar gain, wind speed, and thermal current effects throughout the ECP will essentially be at equilibrium conditions under initial stagnant conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 24-hour Frequency is based on operating experience related to the trending of the ECP temperature during the applicable MODES. This SR verifies that the ECP average water temperature at the point of discharge from the ECP (i.e., SWS suction) is  100 &deg;F.
This SR is modified by a Note indicating that the temperature monitoring is required to be performed only during the summer months (i.e., June 1 to September 30). During other periods of the year, the ECP temperature will not have the potential to reach the temperature limit.
SR 3.7.8.3 This SR (together with SR 3.7.8.1 and 3.7.8.4) verifies that adequate inventory exists to support long term (30 days) cooling. Soundings are performed to ensure the water volume is within limits and that the indicated water level is indicative of an equivalent water volume for accident mitigation. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12-month Frequency reflects the gradual pace of degradation of the physical properties of the ECP.
ANO-1                                          B 3.7.8-3                        Amendment No. 215 Rev. 13,15,
 
ECP B 3.7.8 SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.8.4 This SR (together with SR 3.7.8.1 and 3.7.8.3) verifies that adequate inventory exists to support long term (30 days) cooling. Visual inspections of the loose stone (riprap) placed on the banks of the ECP and of the concrete spillway are performed to ensure erosion, undercut caused by wave action, or any physical degradation is within acceptable limits to enable the ECP to fulfill its safety function. An engineering evaluation of any apparent changes in visual appearance or other abnormal degradation is performed within 7 days to determine OPERABILITY. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 12-month Frequency reflects the gradual pace of degradation of the physical properties of the ECP.
REFERENCES
: 1. SAR, Section 9.3.
: 2. Regulatory Guide 1.27, Rev. 1, Ultimate Heat Sink for Nuclear Power Plants, March 1974.
: 3. 10 CFR 50.36.
ANO-1                                      B 3.7.8-4                        Amendment No. 215 Rev. 15,31,
 
CREVS B 3.7.9 ACTIONS (continued)
F.1 If both CREVS trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable CRE boundary (i.e., Condition B), the CREVS may not be capable of performing the intended function and a loss of safety function has occurred. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.7.9.1 Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not severe, testing each train on a monthly basis adequately checks this system by initiating flow through the HEPA filters and charcoal adsorbers. The CREVS is designed without heaters and need only be operated for 15 minutes to demonstrate the function of the system. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31-day Frequency is based on the known reliability of the equipment and two train redundancy available.
SR 3.7.9.2 This SR verifies that the required CREVS testing is performed in accordance with the 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. Specific test Frequencies and additional information are discussed in detail in the VFTP.
SR 3.7.9.3 This SR verifies that the CREVS automatically isolates the CRE within 10 seconds and switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks on an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months is based on industry operating experience and is consistent with the typical refueling cycle.
SR 3.7.9.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.
ANO-1                                        B 3.7.9-5                      Amendment No. 215 Rev. 34,50,55,
 
CREACS B 3.7.10 ACTIONS (continued)
B.1 and B.2 (continued)
Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
C.1 and C.2 During movement of irradiated fuel, if the Required Action and associated Completion Time of Condition A are not met, the OPERABLE CREACS train must be placed in operation immediately. This action ensures that any active failure will be readily detected.
An alternative to Required Action C.1 is to immediately suspend activities that could release radioactivity that might require the isolation of the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.
D.1 During movement of irradiated fuel assemblies, with two CREACS trains inoperable, action must be taken to immediately suspend activities that could release radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.
E.1 If both CREACS trains are inoperable in MODE 1, 2, 3, or 4, a loss of safety function has occurred, and LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.7.10.1 and SR 3.7.10.2 These SRs, in conjunction with periodic preventative maintenance activities, provide verification that the CREACS will maintain the control room temperature within acceptable bounds. In accordance with SR 3.7.10.1, each train is verified to start and operate for at least 1 hour while maintaining Control Room temperature within the specified limit. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe Frequencies (31 days and 18 months) are appropriate as periodic preventative maintenance activities are routinely performed and significant degradation of the CREACS is not expected over these time periods.
ANO-1                                        B 3.7.10-3                      Amendment No. 215 Rev. 50,55,
 
PRVS B 3.7.11 ACTIONS (continued)
B.1 (continued)
Preplanned measures should be available to address these concerns for intentional and unintentional entry into the Condition. The 24-hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24-hour Completion Time is a typically reasonable time to diagnose, plan and possible repair, and test most problems with the PRVS negative pressure boundary.
C.1 and C.2 If the Required Action and the associated Completion Time are not met, or with both PRVS trains inoperable, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 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 REQUIREMENTS SR 3.7.11.1 Standby systems should be checked periodically to ensure that they function properly. Since the environment and normal operating conditions on this system are not severe, testing each train once a month provides an adequate check on this system. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is based on known reliability of equipment and the two train redundancy available.
SR 3.7.11.2 This SR verifies that the required PRVS testing is performed in accordance with the 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. Specific test frequencies and additional information are discussed in detail in the VFTP.
SR 3.7.11.3 This SR verifies that each PRVS train starts and operates on an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is consistent with the guidance provided in Regulatory Guide 1.52 (Ref. 4).
ANO-1                                      B 3.7.11-3                        Amendment No. 215 Rev. 33,
 
PRVS B 3.7.11 REFERENCES
: 1. SAR, Section 6.5.
: 2. SAR, Sections 14.2.2.5 and 14.2.2.6.
: 3. 10 CFR 50.36.
: 4. Regulatory Guide 1.52, Design, Testing, and Maintenance Criteria for Post Accident Engineered Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water Cooled Nuclear Power Plants, Rev. 2, March 1978.
ANO-1                                  B 3.7.11-4                      Amendment No. 215 Rev.
 
Spent Fuel Pool Water Level B 3.7.13 APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pool since the potential for a release of fission products exists.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
When the initial conditions for an accident cannot be met, immediate action must be taken to preclude the occurrence of an accident. With the spent fuel pool at less than the required level, the movement of fuel assemblies in the spent fuel pool is immediately suspended. This effectively precludes the occurrence of a fuel handling accident. In such a case, unit procedures control the movement of loads over the spent fuel. This does not preclude movement of a fuel assembly to a safe position.
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE REQUIREMENTS SR 3.7.13.1 This SR verifies that sufficient spent fuel pool water is available in the event of a fuel handling accident. The water level in the spent fuel pool must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by unit procedures and are acceptable, based on operating experience.
During refueling operations, the level in the spent fuel pool is at equilibrium with that in the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.
REFERENCES
: 1. FSAR, Section 9.6.1.3.
: 2. FSAR, Section 9.4.
: 3. FSAR, Section 14.2.2.3.
: 4. Regulatory Guide 1.183.
: 5. 10 CFR 50.67.
: 6. 10 CFR 50.36 ANO-1                                        B 3.7.13-2                        Amendment No. 215 Rev. 33,
 
Spent Fuel Pool Boron Concentration B 3.7.14 SURVEILLANCE REQUIREMENTS SR 3.7.14.1 This SR verifies that the concentration of boron in the spent fuel pool and cask loading pit is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over a short period of time.
REFERENCES
: 1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978, NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
: 2. SAR, Section 14.2.2.3.
: 3. Safety Evaluation Report, Section 2.1.3, License Amendment No. 76, April 15, 1983.
: 4. 10 CFR 50.36.
: 5. 10 CFR 50.68.
ANO-1                                        B 3.7.14-3                      Amendment No. 215 Rev. 13,14,
 
AC Sources - Operating B 3.8.1 SURVEILLANCE REQUIREMENTS The AC sources are designed to permit inspection and testing of all important areas and features, especially those that have a standby function, in accordance with SAR, Section 1.4, GDC 18 (Ref. 7). Periodic component tests are supplemented by extensive functional tests during outages (under simulated accident conditions).
Where the SRs discussed herein specify ready-to-load a minimum output voltage of 4000 volts is applicable. This value allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 90% or 3600 V. It also allows for voltage drops to motors and other equipment down through the 120 V level where minimum operating voltage is also usually specified as 90% of name plate rating. The required minimum frequency for loading of the DG is 58.8 Hz (derived from Safety Guide 9); however, this value is not routinely monitored to be within limit within 15 seconds. Meeting minimum frequency is expected prior to the DG voltage reaching the required minimum. This is administratively confirmed on an 18-month interval.
SR 3.8.1.1 This SR ensures correct breaker alignment for each required offsite circuit to ensure that distribution buses and loads are connected to their preferred power source, and that appropriate independence of offsite circuits is maintained. The SR also verifies the indicated availability of three-phase AC electrical power from each required offsite circuit to the onsite distribution network. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7 day Frequency is adequate since breaker position is not likely to change without the operator being aware of it and because its status is displayed in the control room.
SR 3.8.1.2 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR is modified by a Note to indicate that DG starts for this Surveillance may be preceded by an engine prelube period and followed by a warmup period prior to loading.
For the purposes of SR 3.8.1.2 testing with application of the Note, the DGs are started from standby conditions. Standby conditions for a DG means that the diesel engine oil is being continuously circulated and temperature is being maintained consistent with manufacturer recommendations. The signal initiating the start of the DG is varied from one test to another (start with handswitch at control room panel and at DG local control panel) to verify all starting circuits are OPERABLE.
SR 3.8.1.2 requires that the DG starts from standby conditions and achieves ready-to-load conditions (i.e., minimum voltage) within 15 seconds. The 15 second start requirement supports the assumptions of the design basis LOCA analysis in the SAR, Chapter 14 (Ref. 4).
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31-day Frequency provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing.
ANO-1                                      B 3.8.1-12                        Amendment No. 215 Rev. 17,22,50,57,
 
AC Sources - Operating B 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.3 This Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting full rated load. The load test is conducted at 90 to 100 percent of the continuous rating, which is considered to be 90 to 100 percent of the intended service rating, or 2475 kW and  2750 kW. These parameter values contain all necessary allowances for instrument uncertainty. No additional allowances for instrument uncertainty are required to be incorporated in the implementing procedures. A minimum run time of 60 minutes ensures stabilized engine temperatures, while minimizing the time that the DG is connected to the offsite source.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31-day Frequency for this Surveillance provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing.
This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 states that momentary transients (e.g.,
because of changing bus loads) do not invalidate this test. Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
SR 3.8.1.4 This SR provides verification that the level of fuel oil in the engine mounted day tank is being properly maintained. The level is expressed as an equivalent volume in gallons, and is selected to ensure adequate fuel, when combined with the volume contained in one fuel oil storage tank, for not less than 3.5 days operation of one DG loaded to full capacity (Ref. 2).
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is adequate to assure that a sufficient supply of fuel oil is available, since low level alarms are provided and unit operators would be aware of any large uses of fuel oil during this period.
SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day [and engine mounted] tanks once every 31 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and 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 the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 10). This SR is for preventive maintenance. The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during the performance of this Surveillance.
ANO-1                                          B 3.8.1-13                        Amendment No. 215 Rev. 22,50,
 
AC Sources - Operating B 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
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. This 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, and the fuel delivery piping is not obstructed.
The design of the fuel transfer systems is such that pumps operate automatically or must be started manually in order to maintain an adequate volume of fuel oil in the day tanks during DG monthly testing. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramTherefore, a 31-day Frequency is specified to correspond to the interval for DG testing.
SR 3.8.1.7 Transfer of each 4.16 kV ES bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. Reference 1 requires that only one of the two offsite power circuits be capable of automatic transfer. The second (alternate) circuit must be capable of manual transfer, as a minimum. Typically, Startup Transformer No. 1 is aligned for automatic transfer and Startup Transformer No. 2 is aligned to allow manual transfer. In this alignment, the Surveillance verifies the automatic transfer of loads to Startup Transformer No. 1 and the manual transfer of loads to Startup Transformer No. 2. In the event that Startup Transformer No. 2 is aligned for automatic transfer and Startup Transformer No. 1 is aligned for manual transfer, the Surveillance verifies the automatic transfer of loads to Startup Transformer No. 2 and the manual transfer of loads to Startup Transformer No. 1.
For Startup Transformer No. 2, this test also demonstrates the selective load shedding interlock function. (Note: This load shedding function is only required when Startup Transformer No. 2 is selected for automatic transfer.) These features provide protection of required equipment from a sustained degraded grid voltage situation.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency of the Surveillance takes into consideration the unit conditions required to perform the Surveillance (i.e., during refueling shutdown), and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components usually pass the SR when performed at the 18-month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note. The reason for the Note 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, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced.
ANO-1                                        B 3.8.1-14                      Amendment No. 215 Rev. 22,50,
 
AC Sources - Operating B 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.7 (continued)
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODES 1 or 2. Risk insights or deterministic methods may be used for this assessment.
SR 3.8.1.8 This Surveillance demonstrates the as designed operation of the standby power sources during loss of the offsite source. This test verifies all actions encountered from the loss of offsite power, including shedding of the non-essential loads and energization of the emergency buses and respective loads from the DG. It further demonstrates the capability of the DG to automatically achieve ready-to-load conditions (i.e., minimum required voltage) within the specified time.
The DG auto-start time of 15 seconds is derived from requirements of the accident analysis to respond to a design basis large break LOCA. The Surveillance should be continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability has been achieved.
The requirement to verify the connection and power supply of permanent and auto-connected loads, e.g., the running service water pump(s), is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads can not actually be connected or loaded without undue hardship or potential for undesired operation. In lieu of actual demonstration of connection and loading of loads during this test, 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.
If the component start time delays are outside of those assumed by the SAR, component OPERABILITY and DG OPERABILITY must be evaluated.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months takes into consideration unit conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths.
This SR is modified by a Note. The reason for the Note is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.9 In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ES systems so that the fuel, RCS, and reactor building design limits are not exceeded.
ANO-1                                          B 3.8.1-15                      Amendment No. 215 Rev. 22,50,
 
AC Sources - Operating B 3.8.1 SR 3.8.1.9 (continued)
In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ES systems so that the fuel, RCS, and reactor building design limits are not exceeded.
This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.7, during a loss of offsite power actuation test signal in conjunction with an ES actuation signal.
This test is typically conducted by simulating an ESAS signal and either simultaneously or subsequently simulating a LOOP. In certain circumstances, many loads can not actually be connected or loaded without undue hardship or potential for undesired operation. For instance, DHR systems performing a DHR function are not desired to be interrupted from this mode of operation. In lieu of actual demonstration of connection and loading of loads during this test, 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.
For ANO-1, the High Pressure Injection (HPI) pump load is excluded from this DG test; although its auto start logic and load sequencing time are included in the test. HPI pump start under accident flow conditions can challenge LTOP protection considerations when the RCS is closed and challenge pump integrity due to excessive flow or inadequate suction when the RCS is open. The remaining loads that are auto-connected during this test in conjunction with design calculations are sufficient to demonstrate that the DG will perform as designed during a LOCA with coincident LOOP event.
Should the time intervals between two or more loads be reduced such that the interval is less than that assumed in the SAR, the associated DG is conservatively considered to be inoperable unless an evaluation of the condition shows the loading of the DG, with the reduced time interval, to be acceptable. If one or more time delays is inoperable (i.e., the associated component fails to load) or the time interval between two or more loads is greater than assumed in the SAR, then the associated component is considered inoperable, and the appropriate Condition for that component is entered.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 18 months takes into consideration unit conditions required to perform the Surveillance and is intended to be consistent with an expected fuel cycle length of 18 months.
This SR is modified by a Note. The reason for the Note is to minimize wear and tear on the DGs during testing. For the purpose of this testing with application of the Note, the DGs are started from standby conditions, that is, with the engine oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs.
ANO-1                                        B 3.8.1-16                        Amendment No. 215 Rev. 10,22,50,
 
AC Sources - Shutdown B 3.8.2 SURVEILLANCE REQUIREMENTS SR 3.8.2.1 SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, 3, and 4. SR 3.8.1.4 is not required to be met since crediting manual start of the required DG provides sufficiently opportunity to ensure that the fuel oil transfer system is operating properly. SR 3.8.1.7 is not required to be met since only one offsite circuit is required to be OPERABLE. SR 3.8.1.8 and SR 3.8.1.9 are not required to be met because they provide for testing of engineered safeguards actuation signals which are not required to be OPERABLE except in MODES 1, 2, 3, and 4. Automatic actuation and loading of the DGs is not assumed in MODES 5 and 6.
This SR is modified by two Notes. The reason for Note 1 is to preclude requiring the OPERABLE DG from being paralleled with the offsite power network or otherwise rendered inoperable during performance of SRs, and to preclude deenergizing a required 4160 V ES bus or disconnecting a required offsite circuit during performance of this SR. With limited AC sources available, a single event could compromise both the required circuit and the DG. It is the intent that this SR must be capable of being met, but actual performance is not required during periods when the DG and offsite circuit are required to be OPERABLE. When Note 1 is considered, SR 3.8.2.1 requires the following:
SR 3.8.1.1 must be performed and met, SR 3.8.1.2 must be performed and met, SR 3.8.1.3 must be met, but does not have to be performed, SR 3.8.1.4 does not have to be performed or met, SR 3.8.1.5 must be performed and met, SR 3.8.1.6 must be performed and met, SR 3.8.1.7 does not have to be performed or met, SR 3.8.1.8 does not have to be performed or met, and SR 3.8.1.9 does not have to be performed or met.
Note 2 exempts the 15 second start acceptance criteria for SR 3.8.1.2. This allows the AAC DG power source, which does not have auto-start capability or start-time criteria, to be used in lieu of an emergency DG. In MODES 5 and 6, there is sufficient time to manually start a DG in the event the offsite power source is lost. The required DG must be capable of being started from standby conditions and achieving ready-to-load conditions. Although the time to reach ready-to-load conditions is not a part of the acceptance criteria, it is intended that for emergency DG tests, this time be trended to help determine if a condition exists that is degrading the starting capabilities of the DG.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.
ANO-1                                        B 3.8.2-6                    Amendment No. 215,218 Rev. 57,
 
Diesel Fuel Oil and Starting Air B 3.8.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.3.1 (continued)
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31 day Frequency is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and unit operators would be aware of any large uses of fuel oil during this period.
SR 3.8.3.2 The tests of fuel oil prior to addition to the storage tanks are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine operation. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s), but in no case is the time between sampling (and associated results) of new fuel and addition of new fuel oil to the storage tank(s) to exceed 31 days. The tests, limits, and applicable ASTM Standards for the tests listed in Specification 5.5.13, Diesel Fuel Oil Testing Program, are as follows:
: a. Sample the new fuel oil in accordance with ASTM D4057-88 (Ref. 4); and
: b. Verify in accordance with the tests specified in ASTM D975-81 (Ref. 4) that the sample has:
: 1. an absolute specific gravity at 60/60 &deg;F of  0.83 and  0.89 or an API gravity at 60 &deg;F of  27&deg;,  39&deg; (note that vendor-recommended specific gravity limits are normally more restrictive than those listed here and are normally reflected in site procedures),
: 2. a kinematic viscosity at 40 &deg;C of  1.9 centistokes and  4.1 centistokes,
: 3. a flash point of  125 &deg;F, and
: 4. water and sediment within limits.
Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO since the fuel oil is not added to the storage tanks.
Following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975-81 (Ref. 4) are met for new fuel oil when tested in accordance with ASTM D975-81 (Ref. 4), except that the analysis for sulfur may be performed in accordance with ASTM D1552-90 (Ref. 4) or ASTM D2622-87 (Ref. 4). These additional analyses are required by Specification 5.5.13, Diesel Fuel Oil Testing Program, to be performed within 31 days following sampling and addition. This 31 days is intended to assure:
: 1) that the sample taken is not more than 31 days old at the time of adding the fuel oil to the storage tank, and 2) that the results of a new fuel oil sample (sample obtained prior to addition but not more than 31 days prior to) are obtained within 31 days after addition.
ANO-1                                          B 3.8.3-4                        Amendment No. 215 Rev. 18,
 
Diesel Fuel Oil and Starting Air B 3.8.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.3.2 (continued)
For circumstances where multiple fuel oil additions are made within a short period of time, the samples taken for each batch added to the storage tank can be composited for a single follow-up analysis. The 31-day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on DG operation. This Surveillance ensures the availability of high quality fuel oil for the DGs.
Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation. The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.
Particulate concentrations should be determined in accordance with ASTM D2276-88, Method A (Ref. 4). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/l. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing. Each tank is considered and tested separately.
The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.3 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available. The system design requirements provide for a minimum of five engine start cycles without recharging. The pressure specified in this SR is intended to reflect the lowest value at which the five starts can be accomplished.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 31-day Frequency takes into account the capacity, capability, redundancy, and diversity of the AC sources and other indications available in the control room, including alarms, to alert the operator to below normal air start pressure.
SR 3.8.3.4 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks once every 31 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and 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 the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 2). This SR is for preventive maintenance. The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.
ANO-1                                          B 3.8.3-5                        Amendment No. 215 Rev.
 
DC Sources - Operating B 3.8.4 SURVEILLANCE REQUIREMENTS SR 3.8.4.1 Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the battery chargers, which support 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 while supplying the continuous steady-state loads of the associated DC subsystem. On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the minimum float voltage established by the battery manufacturer (2.20 Vpc times the number of connected cells or 127.6 V for a 58 cell battery at the battery terminals). This voltage maintains the battery plates in a condition that supports maintaining the grid life. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7-day Frequency is consistent with manufacturer recommendations.
SR 3.8.4.2 This SR verifies the design capacity of the chargers. According to Regulatory Guide (RG) 1.32 (Ref. 9), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensure that these requirements can be satisfied.
This SR provides two options. One option requires that each battery charger be capable of supplying  300 amps at the minimum established float voltage for 8 hours. The ampere requirements are based on the output rating of the chargers. The voltage requirements are based on the charger voltage level after a response to a loss of AC power. The time period is sufficient for the charger temperature to have stabilized and to have been maintained for at least 2 hours.
The other option requires that each battery charger be capable of recharging the battery after a service test coincident with supplying the largest coincident demands of the various continuous steady state loads (irrespective of the status of the plant during which these demands occur).
This level of loading may not normally be available following the battery service test and will need to be supplemented with additional loads. The duration for this test may be longer than the charger sizing criteria since the battery recharge is affected by float voltage, temperature, and the exponential decay in charging current. The battery is recharged when the measured charging current is  2 amps.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance Frequency is acceptable, given the unit conditions required to perform the test and the other administrative controls existing to ensure adequate charger performance during these 18 month intervals. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths.
ANO-1                                          B 3.8.4-5                        Amendment No. 215 Rev. 49,50,
 
DC Sources - Operating B 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.4.3 A battery service test is a special test of the battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length should correspond to the design duty cycle requirements.
The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance Frequency of 18 months is consistent with the recommendations of RG 1.32 (Ref. 9) and RG 1.129 (Ref. 10), which state that the battery service test should be performed during refueling operations, or at some other outage, with intervals between tests not to exceed 18 months.
This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR.
A modified performance discharge test may be performed in lieu of a service test. The modified performance discharge test (Ref. 8) is a simulated duty cycle consisting of just two rates; the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated one-minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.
A modified performance discharge test is a test of the battery capacity, as found, 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 service test and the test discharge rate must envelope the duty cycle of the service test if the modified performance discharge test is performed in lieu of a service test.
ANO-1                                          B 3.8.4-6                          Amendment No. 215 Rev. 49,50,
 
Battery Parameters B 3.8.6 SURVEILLANCE REQUIREMENTS SR 3.8.6.1 Verifying battery float current while on float charge is used to determine the state of charge of the battery. 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 charged state. The equipment used to monitor float current must have the necessary accuracy and capability to measure electrical currents in the expected range. The float current requirements are based on the float current indicative of a charged battery. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7-day Frequency is consistent with IEEE-450 (Ref. 3).
This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1.
When this float voltage is not maintained the Required Actions of LCO 3.8.4 ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition.
Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.
SR 3.8.6.2 and SR 3.8.6.5 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 127.6 V at the battery terminals, or 2.20 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable. Float voltages in this range or less, but greater than 2.07 Vpc, are addressed in Specification 5.5.6. SRs 3.8.6.2 and 3.8.6.5 require verification that the cell float voltages are equal to or greater than the short term absolute minimum voltage of 2.07 V. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe Frequency for cell voltage verification every 31 days for pilot cell and 92 days for each connected cell is consistent with IEEE-450 (Ref. 3).
SR 3.8.6.3 The limit specified for electrolyte level ensures that the plates suffer no physical damage and maintains adequate electron transfer capability. The minimum design electrolyte level is the minimum level indication mark on the battery cell jar. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with IEEE-450 (Ref. 3).
SR 3.8.6.4 This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60 &deg;F). Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements. Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with IEEE-450 (Ref. 3).
ANO-1                                          B 3.8.6-5                      Amendment No. 215 Rev. 49,
 
Battery Parameters B 3.8.6 SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.6.6 A battery performance discharge test is a test of constant current capacity of a battery after having been in service, to detect any change in the capacity determined by the acceptance test.
The test is intended to determine overall battery degradation due to age and usage.
A battery modified performance discharge test is described in the Bases for SR 3.8.4.3. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6; however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.3.
The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 3), which recommends that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturers 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 manufacturers ratings. Degradation is indicated, according to IEEE-450 (Ref. 3), 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. These Frequencies are consistent with the recommendations in IEEE-450 (Ref. 3).
This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR.
REFERENCES
: 1. SAR, Chapters 8 and 14.
: 2. 10 CFR 50.36.
: 3. IEEE-450-1995, "Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications."
ANO-1                                        B 3.8.6-6                      Amendment No. 215 Rev. 49,
 
Inverters - Operating B 3.8.8 SURVEILLANCE REQUIREMENTS SR 3.8.7.1 This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and 120 VAC buses energized from the inverter. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the 120 VAC buses. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7-day Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter malfunctions.
REFERENCES
: 1. SAR, Chapter 8.
: 2. SAR, Chapter 14.
: 3. 10 CFR 50.36.
: 4. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
ANO-1                                        B 3.8.7-4                          Amendment No. 215 Rev. 13,16,50,
 
Inverters - Shutdown B 3.8.8 SURVEILLANCE REQUIREMENTS SR 3.8.8.1 This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and 120 VAC vital buses energized from the inverter. The verification of proper voltage output ensures that the required power is readily available for the instrumentation connected to the 120 VAC vital buses. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7 day Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter malfunctions.
REFERENCES
: 1. 10 CFR 50.36.
ANO-1                                        B 3.8.8-4                    Amendment No. 215,218 Rev.
 
Distribution Systems - Operating B 3.8.9 SURVEILLANCE REQUIREMENTS SR 3.8.9.1 This Surveillance verifies that the required AC, DC, and 120 VAC bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7-day Frequency takes into account the redundant capability of the AC, DC, and 120 VAC bus electrical power distribution subsystems, and other indications available in the control room that alert the operator to subsystem malfunctions.
REFERENCES
: 1. SAR, Chapter 14.
: 2. 10 CFR 50.36.
: 3. BAW-2441-A, Revision 2, Risk Informed Justification for LCO End-State Changes, September 2006.
ANO-1                                        B 3.8.9-7                        Amendment No. 215 Rev. 1,16,50,
 
Distribution Systems - Shutdown B 3.8.10 ACTIONS (continued)
A.1, A.2.1, A.2.2, A.2.3, A.2.4, A.2.5, and A.2.6 (continued)
Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that which would be required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.
Suspension of these activities does not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of a fuel handling accident. 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 unit safety systems.
Notwithstanding performance of the above conservative Required Actions, a required decay heat removal (DHR) subsystem or a required low temperature overpressure protection (LTOP) feature 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, heat removal and LTOP.
Pursuant to LCO 3.0.6, the DHR ACTIONS and LTOP ACTIONS would not be entered.
Therefore, Required Action A.2.5 is provided to direct declaring DHR inoperable, which results in taking the appropriate DHR actions and Required Action A.2.6 is provided to direct entry into the appropriate LTOP Conditions and Required Actions, which results in taking the appropriate LTOP actions.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the unit safety systems may be without power.
SURVEILLANCE REQUIREMENTS SR 3.8.10.1 This Surveillance verifies that the required AC, DC, and 120 VAC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 7 day Frequency takes into account the capability of the electrical power distribution subsystems, and other indications available in the control room that alert the operator to subsystem malfunctions.
ANO-1                                        B 3.8.10-3                        Amendment No. 215 Rev. 21,
 
Boron Concentration B 3.9.1 ACTIONS (continued)
A.3 (continued)
Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE REQUIREMENTS SR 3.9.1.1 This SR ensures the coolant boron concentration in the RCS and the refueling canal is within the COLR limits. The boron concentration of the coolant in each volume is determined preiodicallyevery 72 hours by chemical analysis. Prior to re-connecting portions of the refueling canal to the RCS, this SR must be met per SR 3.0.4. If any dilution activity has occurred while the cavity was disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is based on industry experience, which has shown 72 hours to be adequate.
REFERENCES
: 1. SAR, Section 1.4, GDC 26.
: 2. 10 CFR 50.36.
ANO-1                                      B 3.9.1-3                        Amendment No. 215 Rev.
 
Nuclear Instrumentation B 3.9.2 SURVEILLANCE REQUIREMENTS SR 3.9.2.1 SR 3.9.2.1 is the performance of a CHANNEL CHECK, which 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 two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. Changes in fuel loading and core geometry can also result in significant differences between source range channels, but each channel should be consistent with its local conditions.
When in MODE 6 with only one channel OPERABLE, a CHANNEL CHECK is still required.
However, in this condition, a redundant source range instrument may not be available for comparison. The CHANNEL CHECK provides verification that the OPERABLE source range channel is energized and indicating a value consistent with current unit status.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified for the same instruments in LCO 3.3.9.
SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION every 18 months. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the source range nuclear instrument is a complete check and re-adjustment of the channel, from the pre-amplifier input to the indicator.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on industry experience which has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
REFERENCES
: 1. SAR, Section 1.4, GDC 13, GDC 26, GDC 28, and GDC 29.
: 2. SAR, Section 14.1.2.4.
: 3. 10 CFR 50.36.
ANO-1                                      B 3.9.2-3                        Amendment No. 215 Rev. 54,
 
Reactor Building Penetrations B 3.9.3 APPLICABILITY The reactor building penetration requirements are applicable during movement of irradiated fuel assemblies within the reactor building because this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, the reactor building penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when movement of irradiated fuel assemblies within the reactor building is not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on reactor building penetration status.
ACTIONS A.1 With the reactor building equipment hatch, air locks, or any reactor building penetration that provides direct access from the reactor building atmosphere to the outside atmosphere not in the required status, the unit must be placed in a condition in which the isolation function is not needed. This is accomplished by immediately suspending movement of irradiated fuel assemblies within the reactor building. Performance of this action shall not preclude moving a component to a safe position.
These actions remove the potential for an event which may require reactor building closure to prevent a significant radioactivity release.
SURVEILLANCE REQUIREMENTS SR 3.9.3.1 This Surveillance demonstrates that each of the reactor building penetrations required to be in its closed position is in that position.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Surveillance is performed every 7 days during the movement of irradiated fuel assemblies within the reactor building. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations.
This Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the reactor building will not result in a release of fission product radioactivity to the environment in excess of that recommended by Standard Review Plan Section 15.7.4 (Ref. 1, 3 and 6).
SR 3.9.3.2 This Surveillance demonstrates that each reactor building isolation valve actuates to its isolation position on manual initiation. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18-month Frequency maintains consistency with other similar reactor building isolation valve testing requirements found in Section 3.6. This ANO-1                                          B 3.9.3-3                          Amendment No. 215 Rev. 33,43,
 
Reactor Building Penetrations B 3.9.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.3.2 (continued)
The SR is modified by a Note stating that this surveillance is not required to be met for valves in isolated penetrations. The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.
SR 3.9.3.3 This SR requires a CHANNEL CALIBRATION of the reactor building purge exhaust radiation monitor. The CHANNEL CALIBRATION is a complete check of the instrument loop and sensor.
The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. The CHANNEL CALIBRATION is performed consistent with the setpoint requirements. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe 18 month Frequency is based on operating experience and is consistent with the typical operating cycle.
REFERENCES
: 1. Safety Evaluation Report related to ANO-1 Amendment No. 195, April 16, 1999.
: 2. SAR, Section 5.2.2.1.3.
: 3. Safety Evaluation Report related to ANO-1 Amendment No. 184, September 20, 1996.
: 4. SAR, Section 14.2.2.3.
: 5. 10 CFR 50.36.
: 6. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.
: 7. Safety Evaluation Report related to ANO-1 Amendment No. 245, August 10, 2011.
ANO-1                                        B 3.9.3-4                        Amendment No. 215 Rev. 33,43,
 
DHR and Coolant Circulation - High Water Level B 3.9.4 ACTIONS (continued)
A.2 If DHR loop requirements are not met, actions shall be taken immediately to suspend the loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core. A minimum refueling canal water level 23 feet above the fuel assemblies seated in the reactor vessel provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading an irradiated fuel assembly, is prudent under this condition.
A.3 If DHR loop requirements are not met, actions shall be initiated immediately in order to satisfy DHR loop requirements.
Restoration of one decay heat removal loop is required because this is the only active method of removing decay heat. Dissipation of decay heat through natural convection to the large inventory of water in the refueling canal should not be relied upon for an extended period of time. The immediate Completion Time reflects the importance of restoring an adequate decay heat removal loop.
A.4 If DHR loop requirements are not met, all reactor building penetrations providing direct access from the reactor building atmosphere to outside atmosphere shall be closed within 4 hours.
If no means of decay heat removal can be restored, the core decay heat could raise temperatures and cause boiling in the core which could result in increased levels of radioactivity in the reactor building atmosphere. Closure of the penetrations providing access to the outside atmosphere will prevent the uncontrolled release of radioactivity to the environment.
SURVEILLANCE REQUIREMENTS SR 3.9.4.1 This Surveillance demonstrates that the DHR loop is in operation and circulating reactor coolant. Verification includes flow, temperature, or pump status monitoring, which help assure that forced flow is providing heat removal. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the DHR System.
ANO-1                                        B 3.9.4-3                      Amendment No. 215 Rev.
 
DHR and Coolant Circulation - Low Water Level B 3.9.5 ACTIONS (continued)
B.1 If no DHR loop is in operation or no DHR loop is OPERABLE, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
B.2 If no DHR loop is in operation or no DHR loop is OPERABLE, actions shall be initiated immediately and continued without interruption to restore one DHR loop to OPERABLE status and operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE DHR loops and one operating DHR loop should be accomplished expeditiously.
If no DHR loop is OPERABLE or in operation, alternate actions shall have been initiated immediately under Condition A to establish  23 ft of water above the top of fuel assemblies seated in the reactor vessel. Furthermore, when the LCO cannot be fulfilled, alternate decay heat removal methods, as specified in the unit's Abnormal and Emergency Operating Procedures, should be implemented. The method used to remove decay heat should be the most prudent as well as the safest choice, based upon unit conditions. The choice could be different if the reactor vessel head is in place rather than removed.
B.3 If no DHR loop is in operation, all reactor building penetrations providing direct access from the reactor building atmosphere to the outside atmosphere must be closed within 4 hours. With the DHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the reactor building atmosphere. Closing reactor building penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded.
The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE REQUIREMENTS SR 3.9.5.1 This Surveillance demonstrates that one DHR loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help assure that forced flow is providing heat removal.
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the DHR system in the control room.
ANO-1                                          B 3.9.5-3                      Amendment No. 215 Rev.
 
DHR and Coolant Circulation - Low Water Level B 3.9.5 SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.5.2 Verification that each required pump is available ensures that an additional DHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to the required pump. Alternatively, verification that a DHR pump is in operation as required by SR 3.9.4.1 also verifies proper breaker alignment and power availability. The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.
REFERENCES
: 1. SAR, Section 1.4.
: 2. SAR, Section 9.5.
: 3. 10 CFR 50.36.
ANO-1                                        B 3.9.5-4                      Amendment No. 215 Rev.
 
Refueling Canal Water Level B 3.9.6 APPLICABILITY LCO 3.9.6 is applicable during movement of irradiated fuel assemblies within the reactor building. The LCO minimizes the possibility of a fuel handling accident in the reactor building that is beyond the assumptions of the safety analysis. If irradiated fuel is not present in the reactor building, there can be no significant radioactivity release as a result of a postulated fuel handling accident in the reactor building.
ACTIONS A.1 With a water level of < 23 feet above the top of the irradiated fuel assemblies seated with the reactor pressure vessel, all operations involving the movement of irradiated fuel assemblies shall be suspended immediately to ensure that a fuel handling accident cannot occur.
The suspension of irradiated fuel movement shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE REQUIREMENTS SR 3.9.6.1 Verification of a minimum water level of 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel limits the consequences of damaged fuel rods that are postulated to result from a postulated fuel handling accident inside the reactor building (Ref. 2).
The periodic Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency of 24 hours is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls, which make significant unplanned level changes unlikely.
REFERENCES
: 1. Regulatory Guide 1.183.
: 2. SAR Section 14.2.2.3.
: 3. 10 CFR 50.67.
: 4. 10 CFR 50.36.
ANO-1                                        B 3.9.6-2                        Amendment No. 215 Rev. 33,
 
ATTACHMENT 6 1CAN031801 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS to 1CAN031801 Page 1 of 2 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATIONS Description of Amendment Request:
The change requests the adoption of an approved change to the standard technical specifications (STS) for Babcock and Wilcox Plants (NUREG-1430), to allow relocation of specific Arkansas Nuclear One - Unit 1 (ANO-1) TS surveillance frequencies to a licensee-controlled program. The proposed change is described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (Rev. 3) (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control - Risk Informed TSTF (RITSTF) Initiative 5b, and was described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).
The proposed changes are consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Traveler, TSTF-425, Rev. 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b. The proposed change relocates 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 5b, Risk-Informed Method for Control of Surveillance Frequencies, (ADAMS Accession No. ML071360456).
Basis for proposed no significant hazards consideration:
As required by 10 CFR 50.91(a), Entergy Operations Inc. (Entergy's) analysis of the issue of no significant hazards consideration is presented below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for periodic surveillance requirements (SRs) to licensee control under a new Surveillance Frequency Control Program (SFPC). Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications (TSs) for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the SRs, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
to 1CAN031801 Page 2 of 2
: 2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed 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.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described 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, Entergy will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based upon the reasoning presented above, Entergy concludes that the requested change does not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of amendment.
 
ATTACHMENT 7 1CAN031801 ANO-1 TO NUREG 1430 SR CROSS-REFERENCE to 1CAN031801 Page 1 of 20 Legend:
Arkansas Nuclear One, Unit 1 (ANO-1) Surveillance Requirement (SR) Frequency Identified for Relocation Not Included in TSTF-425 = Gray Row ANO-1          SR Frequency    SR Frequency ANO-1        NUREG-1430                    ANO-1 Surveillance Description Surveillance        Modified by Modified in Proposed SR #            R4 SR #          {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency          TSTF-425  Amendment Request Section 1.0, Definitions Definition      Definitions                        Staggered Test Basis                              N/A                yes            yes Section 3.1, Reactivity Control Systems Verify SDM [Shutdown Margin] greater than or equal to the limit 3.1.1.1          3.1.1.1                                                                          24 hours              yes            yes specified in the COLR.
Once prior to entering      yes MODE 1 after each                          yes Verify measured core reactivity balance is within +/- 1% k/k of                          (31-EFPD 3.1.2.1          3.1.2.1                                                                        fuel loading                  (31-EFPD Frequency predicted values.                                                                        Frequency AND                                only) 31 EFPD thereafter        only)
Verify individual CONTROL ROD positions are within 6.5% of 3.1.4.1          3.1.4.1                                                                          12 hours              yes            yes their group average height.
Verify CONTROL ROD freedom of movement for each 3.1.4.2          3.1.4.2                                                                          92 days              yes            yes individual CONTROL ROD that is not fully inserted.
3.1.5.1          3.1.5.1    Verify each safety rod is fully withdrawn                              12 hours              yes            yes Verify position of each APSR [Axial Power Shaping Rod] is 3.1.6.1          3.1.6.1                                                                          12 hours              yes            yes within 6.5% of the group average height.
Perform CHANNEL CHECK of required position indicator 3.1.7.1          3.1.7.1                                                                          12 hours              yes            yes channel Perform CHANNEL CALIBRATION of required position 3.1.7.2            N/A                                                                            18 months              no              yes indicator channel.
3.1.8.1          3.1.8.1    Verify THERMAL POWER is  85% RTP                                      1 hour              yes            yes 3.1.8.2          3.1.8.2    Perform SR 3.2.5.1.                                                    2 hours              yes            yes to 1CAN031801 Page 2 of 20 ANO-1          SR Frequency    SR Frequency ANO-1        NUREG-1430                      ANO-1 Surveillance Description Surveillance        Modified by Modified in Proposed SR #          R4 SR #            {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency            TSTF-425  Amendment Request Within 8 hours prior to Verify nuclear overpower trip setpoint  10% RTP higher than performance of 3.1.8.3          3.1.8.3      the THERMAL POWER at which the test is performed, with a                                    yes              no PHYSICS TESTS at maximum setting of 90% RTP.
each test plateau 3.1.8.4          3.1.8.4      Verify SDM to be within the limits provided in the COLR.            24 hours              yes              yes 3.1.9.1          3.1.9.1      Verify THERMAL POWER is  5% RTP                                      1 hour                yes              yes Within 8 hours prior to 3.1.9.2          3.1.9.2      Verify nuclear overpower trip setpoint is  5% RTP.              performance of            yes              no PHYSICS TESTS 3.1.9.3          3.1.9.3      Verify SDM to be within the limit provided in the COLR.              24 hours              yes              yes Section 3.2, Power Distribution Limits Verify regulating rod groups are within the sequence and 3.2.1.1          3.2.1.1                                                                          12 hours              yes              yes overlap limits as specified in the COLR.
Verify regulating rod groups meet the insertion limits as 3.2.1.2          3.2.1.2                                                                          12 hours              yes              yes specified in the COLR.
Verify APSRs are within acceptable limits specified in the 3.2.2.1          3.2.2.1                                                                          12 hours              yes              yes COLR.
Verify AXIAL POWER IMBALANCE is within limits as specified 3.2.3.1          3.2.3.1                                                                          12 hours              yes              yes in the COLR.
7 days AND When QPT has been restored to less than        yes or equal to the steady      (7-day            yes 3.2.4.1          3.2.4.1      Verify QPT is within limits as specified in the COLR state limit, 1 hour for  Frequency  (7-day Frequency only) 12 consecutive hours,        only) or until verified acceptable at  95%
RTP to 1CAN031801 Page 3 of 20 ANO-1        SR Frequency    SR Frequency ANO-1          NUREG-1430                ANO-1 Surveillance Description Surveillance      Modified by Modified in Proposed SR #            R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency        TSTF-425  Amendment Request Section 3.3, Instrumentation 3.3.1.1          3.3.1.1  Perform CHANNEL CHECK                                            12 hours            yes            yes 96 hours AND              yes yes Compare results of calorimetric heat balance calculation to Once within 24 hours    (96-hour 3.3.1.2          3.3.1.2                                                                                                (96-hour Frequency power range channel output                                  after a THERMAL      Frequency POWER change of                          only) only) 10% RTP Compare results of out of core measured AXIAL POWER 3.3.1.3          3.3.1.3                                                                    31 days            yes            yes IMBALANCE to incore measured AXIAL POWER IMBALANCE.
3.3.1.4          3.3.1.4  Perform CHANNEL FUNCTIONAL TEST                                    31 days            yes            yes 3.3.1.5          3.3.1.5  Perform CHANNEL CALIBRATION.                                    18 months            yes            yes N/A            3.3.1.6  {Verify that RPS RESPONSE TIME is within limits}                    N/A              yes              no 3.3.3.1          3.3.3.1  Perform CHANNEL FUNCTIONAL TEST                                    92 days            yes            yes 3.3.4.1          3.3.4.1  Perform CHANNEL FUNCTIONAL TEST                                    92 days            yes            yes 3.3.5.1          3.3.5.1  Perform CHANNEL CHECK                                            12 hours            yes            yes 3.3.5.2          3.3.5.2  Perform CHANNEL FUNCTIONAL TEST                                    31 days            yes            yes 3.3.5.3          3.3.5.3  Perform CHANNEL CALIBRATION                                      18 months            yes            yes N/A            3.3.5.4  {Verify ESFAS RESPONSE TIME within limits}                          N/A              yes              no 3.3.6.1          3.3.6.1  Perform CHANNEL FUNCTIONAL TEST                                  18 months            yes            yes 3.3.7.1          3.3.7.1  Perform digital actuation logic CHANNEL FUNCTIONAL TEST.          31 days            yes            yes 3.3.8.1          3.3.8.1  Perform CHANNEL CHECK                                              7 days            yes            yes N/A            3.3.8.2  {Perform CHANNEL FUNCTIONAL TEST}                                    N/A              yes              no to 1CAN031801 Page 4 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request Perform CHANNEL CALIBRATION with setpoint Allowable Value as follows:
: a. Degraded voltage  423.2 V and  436.0 V with a time 3.3.8.2        3.3.8.3                                                                18 months        yes              yes delay of 8 seconds +/- 1 second; and
: b. Loss of voltage  1600 V and  3000 V with a time delay of 0.30 seconds and  0.98 seconds.
3.3.9.1        3.3.9.1 Perform CHANNEL CHECK                                          12 hours        yes              yes 3.3.9.2        3.3.9.2 Perform CHANNEL CALIBRATION                                    18 months        yes              yes 3.3.10.1      3.3.10.1 Perform CHANNEL CHECK                                          12 hours        yes              yes 3.3.10.2        N/A  Perform CHANNEL FUNCTIONAL TEST                                31 days        no              yes 3.3.10.3      3.3.10.2 Perform CHANNEL CALIBRATION                                    18 months        yes              yes 3.3.11.1      3.3.11.1 Perform CHANNEL CHECK                                          12 hours        yes              yes 3.3.11.2      3.3.11.2 Perform CHANNEL FUNCTIONAL TEST                                31 days        yes              yes 3.3.11.3      3.3.11.3 Perform CHANNEL CALIBRATION                                    18 months        yes              yes N/A        3.3.11.4 {Verify EFIC RESPONSE TIME is within limits}                      N/A          yes              no 3.3.12.1      3.3.12.1 Perform CHANNEL FUNCTIONAL TEST                                31 days        yes              yes 3.3.13.1      3.3.13.1 Perform CHANNEL FUNCTIONAL TEST                                31 days        yes              yes 3.3.14.1      3.3.14.1 Perform CHANNEL FUNCTIONAL TEST                                31 days        yes              yes N/A        3.3.15.1 {Perform CHANNEL CHECK}                                          N/A          yes              no N/A        3.3.15.2 {Perform CHANNEL FUNCTIONAL TEST}                                N/A          yes              no
{Perform CHANNEL CALIBRATION with setpoint Allowable N/A        3.3.15.3                                                                  N/A          yes              no Value  [25] mR/hr}
Perform CHANNEL CHECK for each required instrumentation 3.3.15.1      3.3.17.1                                                                31 days        yes              yes channel that is normally energized.
3.3.15.2      3.3.17.2 Perform CHANNEL CALIBRATION.                                  18 months        yes              yes 3.3.16.1      3.3.16.1 Perform CHANNEL CHECK                                          12 hours        yes              yes to 1CAN031801 Page 5 of 20 ANO-1    SR Frequency  SR Frequency ANO-1        NUREG-1430                        ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #          R4 SR #            {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request 3.3.16.2        3.3.16.2    Perform CHANNEL FUNCTIONAL TEST                                    31 days        yes              yes 3.3.16.3        3.3.16.3    Perform CHANNEL CALIBRATION                                      18 months        yes              yes
{[Perform CHANNEL CHECK for each required instrumentation N/A          3.3.18.1                                                                          N/A          yes              no channel that is normally energized]}
{Verify each required control circuit and transfer switch is N/A          3.3.18.2                                                                          N/A          yes              no capable of performing the intended function.}
{Perform CHANNEL CALIBRATION for each required N/A          3.3.18.3                                                                          N/A          yes              no instrumentation channel.}
Section 3.4, Reactor Coolant System Verify RCS loop pressure is within the limit specified in the 3.4.1.1          3.4.1.1                                                                        12 hours        yes              yes COLR.
Verify RCS hot leg temperature is within the limit specified in 3.4.1.2          3.4.1.2                                                                        12 hours        yes              yes the COLR.
3.4.1.3          3.4.1.3    Verify RCS total flow is within the limit specified in the COLR. 12 hours        yes              yes Verify RCS total flow rate is within the limit specified in the 3.4.1.4          3.4.1.4                                                                      18 months        yes              yes COLR by measurement.
3.4.2.1          3.4.2.1    Verify RCS Tavg  525 &deg;F.                                          12 hours        yes              yes Verify RCS pressure, RCS temperature, and RCS heatup rates 3.4.3.1          3.4.3.1                                                                      30 minutes      yes              yes are within the limits specified in Figure 3.4.3-1.
Verify RCS pressure, RCS temperature, and RCS cooldown 3.4.3.2            N/A                                                                        30 minutes      no              yes rates are within the limits specified in Figure 3.4.3-2.
Verify RCS pressure, RCS temperature, and RCS cooldown 3.4.3.3            N/A                                                                        30 minutes      no              yes rates are within the limits specified in Figure 3.4.3-3.
Verify RCS pressure and RCS temperature are within the 3.4.3.4            N/A                                                                        30 minutes      no              yes criticality limits specified in Figure 3.4.3-1.
3.4.4.1          3.4.4.1    Verify required RCS loops are in operation.                        12 hours        yes              yes 3.4.5.1          3.4.5.1    Verify required RCS loop is in operation.                          12 hours        yes              yes to 1CAN031801 Page 6 of 20 ANO-1        SR Frequency    SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance    Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency        TSTF-425  Amendment Request Verify correct breaker alignment and indicated power available 3.4.5.2        3.4.5.2                                                                      7 days            yes            yes to each required pump.
3.4.6.1        3.4.6.1 Verify required DHR or RCS loop is in operation                    12 hours            yes            yes Verify correct breaker alignment and indicated power available 3.4.6.2        3.4.6.2                                                                      7 days            yes            yes to each required pump 3.4.7.1        3.4.7.1 Verify required DHR loop is in operation                            12 hours            yes            yes 3.4.7.2        3.4.7.2 Verify required SG secondary side water levels are  20 inches. 12 hours            yes            yes Verify correct breaker alignment and indicated power available 3.4.7.3        3.4.7.3                                                                      7 days            yes            yes to each required DHR pump.
3.4.8.1        3.4.8.1 Verify required DHR loop is in operation                            12 hours            yes            yes Verify correct breaker alignment and indicated power available 3.4.8.2        3.4.8.2                                                                      7 days            yes            yes to each required DHR pump.
3.4.9.1        3.4.9.1 Verify pressurizer water level  320 inches.                        12 hours            yes            yes 3.4.9.2        3.4.9.2 Verify capacity of ES bus powered pressurizer heaters  126 kW. 18 months            yes            yes
{[Verify emergency power supply for pressurizer heaters is N/A          3.4.9.3                                                                      N/A              yes              no OPERABLE]}
N/A        3.4.11.1 {Perform one complete cycle of the block valve}                      N/A              yes              no N/A        3.4.11.2 {Perform one complete cycle of the PORV}                              N/A              yes              no
{[Verify PORV and block valve are capable of being powered N/A        3.4.11.3                                                                      N/A              yes              no from an emergency power source.]}
{Verify a maximum of [one] makeup pump is capable of N/A        3.4.12.1                                                                      N/A              yes              no injecting into the RCS}
30 minutes during      yes RCS heatup and                        yes Verify pressurizer level does not represent a water solid                            (12-hour 3.4.11.1      3.4.12.4                                                                    cooldown                    (12-hour Frequency condition                                                                          Frequency AND                              only) 12 hours          only) 3.4.11.2      3.4.12.2 Verify HPI is deactivated                                          12 hours            yes            yes to 1CAN031801 Page 7 of 20 ANO-1        SR Frequency  SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance      Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency        TSTF-425  Amendment Request 3.4.11.3      3.4.12.3 Verify each pressurized CFT is isolated.                                    12 hours            yes              yes 3.4.12.5 3.4.11.4              Verify OPERABLE pressure relief capability                                  12 hours            yes              yes 3.4.12.6 N/A        3.4.12.7 {Perform CHANNEL FUNCTIONAL TEST for PORV}                                    N/A              yes              no 3.4.11.5      3.4.12.8 Perform CHANNEL CALIBRATION of ERV opening circuitry.                      18 months            yes              yes Verify reactor coolant DOSE EQUIVALENT XE-133 specific 3.4.12.1      3.4.16.1                                                                              7 days            yes              yes activity  2200 Ci/gm.
Verify reactor coolant DOSE EQUIVALENT I-131 specific 3.4.12.2      3.4.16.2                                                                            14 days            yes              yes activity  1.0 Ci/gm.
N/A        3.4.16.3 {Determine E bar}                                                              N/A              yes              no Verify RCS operational LEAKAGE is within limits by 3.4.13.1      3.4.13.1                                                                            72 hours            yes              yes performance of an RCS water inventory balance.
Verify primary to secondary LEAKAGE is  150 gallons per day 3.4.13.2      3.4.13.2                                                                            72 hours            yes              yes through any one SG.
INSERVICE TESTING PROGRAM AND Once prior to entering Verify leakage from each RCS pressure isolation check valve, MODE 2 whenever or pair of check valves, as applicable, is less than or equal to an 3.4.14.1      3.4.14.1                                                                      the unit has been in      yes              no equivalent of the Allowable Leakage Limit identified below at a MODE 5 for 7 days or differential test pressure  150 psid.
more, if leakage testing has not been performed in the previous 9 months Verify DHR System autoclosure interlock prevents the valves 3.4.14.2      3.4.14.2 from being opened with a simulated or actual high RCS                      18 months            yes              yes pressure signal.
to 1CAN031801 Page 8 of 20 ANO-1    SR Frequency  SR Frequency ANO-1        NUREG-1430                    ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #          R4 SR #        {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request Verify DHR System autoclosure interlock causes the valves to close automatically with a simulated or actual high RCS 3.4.14.3        3.4.14.3  pressure signal:                                                18 months        yes              yes
: a. 340 psig for one valve; and
: b. 400 psig for the other valve.
Verify DHR System autoclosure interlock prevents the valves 3.4.14.4          N/A      from being opened with a simulated or actual Core Flood Tank    18 months        no              yes isolation valve not closed signal.
Verify DHR System autoclosure interlock causes the valves to 3.4.14.5          N/A      close automatically with a simulated or actual Core Flood Tank  18 months        no              yes isolation valve not closed signal.
Perform CHANNEL CHECK of required reactor building 3.4.15.1        3.4.15.1                                                                    12 hours        yes              yes atmosphere radioactivity monitor.
Perform CHANNEL FUNCTIONAL TEST of required reactor 3.4.15.2        3.4.15.2                                                                    92 days        yes              yes building atmosphere radioactivity monitor Perform CHANNEL CALIBRATION of required reactor building 3.4.15.3        3.4.15.4                                                                  18 months        yes              yes atmosphere radioactivity monitor.
Perform CHANNEL CALIBRATION of required reactor building 3.4.15.4        3.4.15.3                                                                  18 months        yes              yes sump monitor.
Section 3.5, Emergency Core Cooling Systems (ECCS) 3.5.1.1        3.5.1.1    Verify each CFT isolation valve is fully open                    12 hours        yes              yes 3
Verify borated water volume in each CFT is  970 ft and 3.5.1.2          3.5.1.2            3                                                      12 hours        yes              yes 1110 ft .
Verify nitrogen cover pressure in each CFT is  560 psig and 3.5.1.3          3.5.1.3                                                                    12 hours        yes              yes 640 psig.
to 1CAN031801 Page 9 of 20 ANO-1          SR Frequency    SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance        Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency          TSTF-425  Amendment Request 31 days AND Once within 12 hours after each solution        yes level increase of                          yes 3.5.1.4      3.5.1.4  Verify boron concentration in each CFT is  2270 ppm.                0.2 feet that is not    (31-day (31-day Frequency the result of addition    Frequency only) from a borated water        only) source of known concentration 2270 ppm Verify power is removed from each CFT isolation valve 3.5.1.5      3.5.1.5                                                                              31 days              yes            yes operator.
{[Verify the following valves are in the listed position with power N/A        3.5.2.1                                                                                N/A                yes              no to the valve operator removed.]}
Verify each ECCS manual, power operated, and automatic 3.5.2.1      3.5.2.2  valve in the flow path, that is not locked, sealed, or otherwise            31 days              yes            yes secured in position, is in the correct position.
N/A        3.5.2.3  {[Verify ECCS piping is full of water.]}                                      N/A                yes              no Verify each ECCS automatic valve in the flow path that is not 3.5.2.3      3.5.2.5  locked, sealed, or otherwise secured in position, actuates to the          18 months              yes            yes correct position on an actual or simulated actuation signal.
Verify each ECCS pump starts automatically on an actual or 3.5.2.4      3.5.2.6                                                                            18 months              yes            yes simulated actuation signal.
{[Verify the correct settings of stops for the following HPI stop N/A        3.5.2.7                                                                                N/A                yes              no check valves:]}
{[Verify the flow controllers for the following LPI throttle valves N/A        3.5.2.8                                                                                N/A                yes              no operate properly:]}
Verify, by visual inspection, each ECCS train reactor building 3.5.2.5      3.5.2.9  sump suction inlet is not restricted by debris and screens show            18 months              yes            yes no evidence of structural distress or abnormal corrosion.
3.5.4.1      3.5.4.1  Verify BWST borated water temperature is  40 &deg;F and  110 &deg;F.              24 hours              yes            yes to 1CAN031801 Page 10 of 20 ANO-1    SR Frequency  SR Frequency ANO-1        NUREG-1430                      ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #          R4 SR #          {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request 3.5.4.2          3.5.4.2    Verify BWST borated water level is  38.4 feet and  42 feet.            7 days        yes              yes 3.5.4.3          3.5.4.3    Verify BWST boron concentration  2270 ppm and  2670 ppm.              7 days        yes              yes Section 3.6, Reactor Building Systems 3.6.2.2          3.6.2.2    Verify only one door in the air lock can be opened at a time.          18 months        yes              yes 3.6.3.1 3.6.3.1                      Verify each reactor building purge isolation valve is closed.          31 days        yes              yes 3.6.3.2 Verify each reactor building isolation manual valve and blind flange that is located outside the reactor building and not locked, sealed, or otherwise secured, and is required to be 3.6.3.2          3.6.3.3                                                                            31 days        yes              yes closed during accident conditions is closed, except for reactor building isolation valves that are open under administrative controls.
Verify the isolation time of each automatic power operated            INSERVICE 3.6.3.4          3.6.3.5                                                                                            yes              no reactor building isolation valve is within limits.                TESTING PROGRAM
{Perform leakage rate testing for containment purge valves with N/A            3.6.3.6                                                                              N/A          yes              no resilient seals}
Verify each automatic reactor building isolation valve that is not 3.6.3.5          3.6.3.7    locked, sealed, or otherwise secured in position, actuates to the      18 months        yes              yes isolation position on an actual or simulated actuation signal.
{[Verify each [ ] inch containment purge valve is blocked to N/A            3.6.3.8                                                                              N/A          yes              no restrict the valve from opening > [50]%.]}
3.6.4.1          3.6.4.1    Verify reactor building pressure is  -1.0 psig and  +3.0 psig.        12 hours        yes              yes N/A            3.6.5.1    {Verify containment average air temperature is within limit.}            N/A          yes              no Verify each reactor building spray manual, power operated, and automatic valve in each required flow path that is not locked, 3.6.5.1          3.6.6.1                                                                            31 days        yes              yes sealed, or otherwise secured in position is in the correct position.
Operate each required reactor building cooling train fan unit for 3.6.5.2          3.6.6.2                                                                            31 days        yes              yes 15 minutes.
to 1CAN031801 Page 11 of 20 ANO-1          SR Frequency  SR Frequency ANO-1        NUREG-1430                  ANO-1 Surveillance Description Surveillance        Modified by Modified in Proposed SR #            R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency          TSTF-425  Amendment Request Verify each required reactor building cooling train cooling water 3.6.5.3          3.6.6.3                                                                          31 days              yes              yes flow rate is  1200 gpm.
Verify each automatic reactor building spray valve in each required flow path that is not locked, sealed, or otherwise 3.6.5.5          3.6.6.5                                                                        18 months              yes              yes secured in position, actuates to the correct position on an actual or simulated actuation signal.
Verify each required reactor building spray pump starts 3.6.5.6          3.6.6.6                                                                        18 months              yes              yes automatically on an actual or simulated actuation signal.
Verify each required reactor building cooling train starts 3.6.5.7          3.6.6.7                                                                        18 months              yes              yes automatically on an actual or simulated actuation signal.
Following maintenance which 3.6.5.8          3.6.6.8 Verify each spray nozzle is unobstructed.                                                      yes              no could result in nozzle blockage Verify each Spray Additive System manual, power operated, 3.6.6.1          3.6.7.1 and automatic valve in the flow path that is not locked, sealed,          31 days              yes              yes or otherwise secured in position is in the correct position.
3.6.6.2          3.6.7.2 Verify sodium hydroxide tank solution volume is  9000 gallons.          184 days              yes              yes Verify sodium hydroxide tank solution concentration is 3.6.6.3          3.6.7.3                                                                          184 days              yes              yes
                              > 6.0 wt% and < 8.5 wt.% NaOH.
Verify each Spray Additive System automatic valve in the flow 3.6.6.4          3.6.7.4 path actuates to the correct position on an actual or simulated        18 months              yes              yes actuation signal.
{Verify Spray Additive System flow [rate] from each solution's N/A            3.6.7.5                                                                            N/A                yes              no flow path.}
Section 3.7, Plant Systems Verify each MSIV actuates to the isolation position on an actual 3.7.2.2          3.7.2.2                                                                        18 months              yes              yes or simulated actuation signal.
to 1CAN031801 Page 12 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request Verify that each MFIV, Main Feedwater Block Valve, Low Load Feedwater Control Valve and Startup Feedwater Control Valve 3.7.3.2        3.7.3.2                                                                    18 months        yes              yes actuates to the isolation position on an actual or simulated actuation signal.
{Verify one complete cycle of each AVV [Atmospheric Vent N/A          3.7.4.1                                                                        N/A          yes              no Valve].}
N/A          3.7.4.2 {Verify one complete cycle of each AVV block valve.}                  N/A          yes              no Verify the specific activity of the secondary coolant is 3.7.4.1      3.7.17.1                                                                      31 days        yes              yes 0.1 &#xb5;Ci/gm DOSE EQUIVALENT I-131.
Verify each EFW manual, power operated, and automatic valve in each water flow path and in both steam supply flow paths to 3.7.5.1        3.7.5.1                                                                      31 days        yes              yes the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.
Verify each EFW automatic valve that is not locked, sealed, or 3.7.5.3        3.7.5.3 otherwise secured in position, actuates to the correct position    18 months        yes              yes on an actual or simulated actuation signal.
Verify each EFW pump starts automatically on an actual or 3.7.5.4        3.7.5.4                                                                    18 months        yes              yes simulated actuation signal.
Verify that feedwater is delivered to each steam generator using 3.7.5.6          N/A                                                                      18 months        no              yes the motor-driven EFW pump.
{[ Perform a CHANNEL FUNCTIONAL TEST for the EFW pump N/A          3.7.5.6                                                                        N/A          yes              no suction pressure interlocks.]}
{[ Perform a CHANNEL CALIBRATION for the EFW pump N/A          3.7.5.7                                                                        N/A          yes              no suction pressure interlocks.]}
Verify QCST volume is  267,000 gallons when required for 3.7.6.1        3.7.6.1                                                                      12 hours        yes              yes both units and  107,000 gallons when only required for Unit 1.
{Verify each CCW manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is N/A          3.7.7.1                                                                        N/A          yes              no not locked, sealed, or otherwise secured in position, is in the correct position.}
to 1CAN031801 Page 13 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request
{Verify each CCW automatic valve in the flow path that is not N/A          3.7.7.2 locked, sealed, or otherwise secured in position, actuates to the    N/A          yes              no correct position on an actual or simulated actuation signal.}
{Verify each CCW pump starts automatically on an actual or N/A          3.7.7.3                                                                      N/A          yes              no simulated actuation signal.}
Verify each SWS manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not 3.7.7.1        3.7.8.1                                                                    31 days        yes              yes locked, sealed, or otherwise secured in position, is in the correct position.
Verify each SWS automatic valve in the flow path that is not 3.7.7.2        3.7.8.2 locked, sealed, or otherwise secured in position, actuates to the  18 months        yes              yes correct position on an actual or simulated actuation signal.
Verify each required SWS pump starts automatically on an 3.7.7.3        3.7.8.3                                                                    18 months        yes              yes actual or simulated signal.
Verify that the indicated water level of the ECP [Emergency 3.7.8.1        3.7.9.1 Cooling Pond] is greater than or equal to that required for an      24 hours        yes              yes ECP volume of 70 acre-ft.
3.7.8.2      3.7.9.2  Verify average water temperature is  100 &deg;F.                      24 hours        yes              yes N/A          3.7.9.3 {[Operate each cooling tower fan for> [15) minutes.]}                N/A          yes              no Perform soundings of the ECP to verify:
: 1. A contained water volume of ECP  70 acre-feet, and 3.7.8.3          N/A                                                                      12 months        no              yes
: 2. The minimum indicated water level needed to ensure a volume of 70 acre-feet is maintained.
Perform visual inspection of the ECP to verify conformance with 3.7.8.4          N/A                                                                      12 months        no              yes design requirements.
Operate each CREVS [Control Room Emergency Ventilation 3.7.9.1      3.7.10.1                                                                    31 days        yes              yes System] train for  15 minutes Verify the CREVS automatically isolates the Control Room and 3.7.9.3      3.7.10.3 switches into a recirculation mode of operation on an actual or    18 months        yes              yes simulated actuation signal to 1CAN031801 Page 14 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request
[Verify the system makeup flow rate is  [270] and  [330] cfm N/A        3.7.10.5                                                                        N/A          yes              no when supplying the the control room with outside air.]}
Verify each CREACS [Control Room Emergency Air 3.7.10.1      3.7.11.1 Conditioning System] train starts, operates for at least 1 hour,    31 days        yes              yes and maintains control room air temperature  84 &deg;F D. B.
3.7.10.2        N/A    Verify system flow rate of 9900 cfm +/- 10%.                          18 months        no              yes Operate each PRVS [Penetration Room Ventilation System]
3.7.11.1      3.7.12.1                                                                      31 days        yes              yes train for  15 minutes.
Verify each PRVS train actuates on an actual or simulated 3.7.11.3      3.7.12.3                                                                    18 months        yes              yes actuation signal.
{Verify one EVS train can maintain a pressure  [ ] inches water N/A        3.7.12.4 gauge relative to atmospheric pressure during the [post                N/A          yes              no accident] mode of operation at a flow rate of  [3000) cfm.}
{[Verify each EVS filter cooling bypass damper can be N/A        3.7.12.5                                                                        N/A          yes              no opened.]}
{[Operate each FSPVS train for [ 10 continuous hours with the N/A        3.7.13.1 heaters operating or (for systems without heaters)                    N/A          yes              no 15 minutes].}
{[Perform required FSPVS filter testing in accordance with the N/A        3.7.13.2                                                                        N/A          yes              no
[Ventilation Filter Testing Program (VFTP)].}
{[Verify each FSPVS train actuates on an actual or simulated N/A        3.7.13.3                                                                        N/A          yes              no actuation signal.]}
{Verify one FSPVS train can maintain a pressure  [ ] inches N/A        3.7.13.4 water gauge with respect to atmospheric pressure during the            N/A          yes              no
[post accident] mode of operation at a flow rate  [3000] cfm.}
N/A        3.7.13.5 {[Verify each FSPVS filter bypass damper can be opened.]}              N/A          yes              no Verify the spent fuel pool water level is  23 ft above the top of 3.7.13.1      3.7.14.1                                                                      7 days        yes              yes irradiated fuel assemblies seated in the storage racks.
3.7.14.1      3.7.15.1 Verify the spent fuel pool boron concentration is > 2000 ppm.        7 days        yes              yes N/A        3.7.18.1 {Verify steam generator water level to be within limits.}              N/A          yes              no to 1CAN031801 Page 15 of 20 ANO-1    SR Frequency  SR Frequency ANO-1          NUREG-1430                      ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #            R4 SR #          {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request Section 3.8, Electrical Power Systems Verify correct breaker alignment and indicated power availability 3.8.1.1            3.8.1.1                                                                          7 days        yes              yes for each required offsite circuit 3.8.1.2    Verify each DG starts from standby conditions and, in 3.8.1.2                                                                                            31 days        yes              yes 3.8.1.7    15 seconds achieves ready-to-load conditions Verify each DG is synchronized and loaded and operates for 3.8.1.3            3.8.1.3                                                                        31 days        yes              yes 60 minutes at a load  2475 kW and  2750 kW.
3.8.1.4            3.8.1.4    Verify each day tank contains  160 gallons of fuel oil              31 days        yes              yes 3.8.1.5            3.8.1.5    Check for and remove accumulated water from each day tank.          31 days        yes              yes Verify the fuel oil transfer system operates to transfer fuel oil 3.8.1.6            3.8.1.6                                                                        31 days        yes              yes from storage tanks to the day tank.
Verify automatic transfer of AC power sources to the selected 3.8.1.7            3.8.1.8    offsite circuit and manual transfer to the alternate required      18 months        yes              yes offsite circuit.
{Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and:
: a. Following load rejection, the frequency is  [63] Hz, N/A              3.8.1.9    b. Within [3] seconds following load rejection, the voltage is      N/A          yes              no
[3740] V and  [4580} V, and
: c. Within [3} seconds following load rejection, the frequency is
[58.8] Hz and  [61.2] Hz.}
{Verify each DG does not trip, and voltage is maintained N/A            3.8.1.10    [5000] V during and following a load rejection of  [4500] kW        N/A          yes              no and  [5000] kW.}
to 1CAN031801 Page 16 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request Verify on an actual or simulated loss of offsite power signal:
: a. De-energization of emergency buses;
: b. Load shedding from emergency buses; and
: c. DG auto-starts from standby condition and:
3.8.1.8      3.8.1.11      1. achieves ready-to-load conditions in  15 seconds,    18 months        yes              yes
: 2. energizes permanently connected loads,
: 3. energizes auto-connected shutdown load through automatic load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
{Verify on an actual or simulated [Engineered Safety Feature (ESF)] actuation signal each DG auto-starts from standby condition and:
: a. In  [12] seconds after auto-start and during tests, achieves voltage  [37 40] V and frequency  [58.8] Hz,
: b. Achieves steady state voltage  [37 40] V and  [4580] V N/A        3.8.1.12      and frequency  [58.8] Hz and  [61.2] Hz,                        N/A          yes              no
: c. Operates for  5 minutes,
: d. Permanently connected loads remain energized from the offsite power system, and
: e. Emergency loads are energized [or autoconnected through the automatic load sequencer] from the offsite power system.}
{Verify each DG's noncritical automatic trips are bypassed on
[actual or simulated loss of voltage signal on the emergency N/A        3.8.1.13                                                                        N/A          yes              no bus concurrent with an actual or simulated ESF actuation signal].}
{Verify each DG operates for  24 hours:
: a. For  [2] hours loaded  [5250] kW and  [6000] kW and N/A        3.8.1.14                                                                        N/A          yes              no
: b. For the remaining hours of the test loaded  [4500] kW and
[5000] kW.}
to 1CAN031801 Page 17 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request
{Verify each DG starts and achieves:
: a. In  [10] seconds, voltage  [3740] V and frequency N/A        3.8.1.15      [58.8] Hz and                                                N/A          yes              no
: b. Steady state voltage  [3740] V and  [4580] V, and frequency  [58.8] Hz and  [61.2] Hz.}
{Verify each DG:
: a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite N/A        3.8.1.16      power,                                                        N/A          yes              no
: b. Transfers loads to offsite power source, and
: c. Returns to ready-to-load operation.}
{Verify, with a DG operating in test mode and connected to its bus, an actual or simulated ESF actuation signal overrides the test mode by:
N/A        3.8.1.17                                                                    N/A          yes              no
: a. Returning DG to ready-to-load operation and
[b. Automatically energizing the emergency load from offsite power.]}
{Verify interval between each sequenced load block is within N/A        3.8.1.18 +/- [10% of design interval] for each emergency [and shutdown]      N/A          yes              no load sequencer}
Verify on an actual or simulated loss of offsite power in conjunction with an actual or simulated ESF actuation signal:
: a. De-energization of emergency buses;
: b. Load shedding from emergency buses; and
: c. DG auto-starts from standby condition and:
3.8.1.9      3.8.1.19                                                                18 months        yes              yes
: 1. achieves ready-to-load conditions in  15 seconds,
: 2. energizes permanently connected loads,
: 3. energizes auto-connected shutdown load through automatic load sequencing timers, and
: 4. supplies connected loads for  5 minutes.
to 1CAN031801 Page 18 of 20 ANO-1    SR Frequency  SR Frequency ANO-1      NUREG-1430                  ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #        R4 SR #      {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency    TSTF-425  Amendment Request
{Verify, when started simultaneously from standby condition, N/A        3.8.1.20 each DG achieves, in  [10] seconds, voltage  [3740) V and            N/A          yes              no
[4580) V, and frequency  [58.8] Hz and  [61.2] Hz.}
For AC Sources required to be OPERABLE, the SRs of Specification 3.8.1, "AC Sources - Operating," except 3.8.2.1        3.8.2.1                                                                      31 days        no              yes SR 3.8.1.4, SR 3.8.1.7, SR 3.8.1.8, and SR 3.8.1.9, are applicable.
Verify each fuel oil storage tank contains  20,000 gallons of 3.8.3.1        3.8.3.1                                                                      31 days        yes              yes fuel.
N/A          3.8.3.2 {Verify lube oil inventory is  a [7] day supply}                      N/A          yes              no 3.8.3.3        3.8.3.4 Verify each DG required air start receiver pressure is  175 psig. 31 days        yes              yes Check for and remove accumulated water from each fuel oil 3.8.3.4        3.8.3.5                                                                      31 days        yes              yes storage tank.
Verify battery terminal voltage is greater than or equal to the 3.8.4.1        3.8.4.1                                                                      7 days        yes              yes minimum established float voltage.
Verify each battery charger supplies  300 amps at greater than or equal to the minimum established float voltage for  8 hours.
OR 3.8.4.2        3.8.4.2 Verify each battery charger can recharge the battery to the fully  18 months        yes              yes charged state within 24 hours while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the 3.8.4.3        3.8.4.3                                                                    18 months        yes              yes design duty cycle when subjected to a battery service test or a modified performance discharge test 3.8.6.1        3.8.6.1 Verify each battery float current is  2 amps                        7 days        yes              yes 3.8.6.2        3.8.6.2 Verify each battery pilot cell float voltage is  2.07 V            31 days        yes              yes Verify each battery connected cell electrolyte level is greater 3.8.6.3        3.8.6.3                                                                      31 days        yes              yes than or equal to minimum established design limits.
to 1CAN031801 Page 19 of 20 ANO-1        SR Frequency    SR Frequency ANO-1        NUREG-1430                    ANO-1 Surveillance Description Surveillance      Modified by Modified in Proposed SR #            R4 SR #        {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency        TSTF-425  Amendment Request Verify each battery pilot cell temperature is greater than or 3.8.6.4          3.8.6.4                                                                            31 days            yes            yes equal to minimum established design limits 3.8.6.5          3.8.6.5    Verify each battery connected cell float voltage is  2.07 V.          92 days            yes            yes 60 months AND 12 months when battery shows degradation, or has reached 85% of the expected life with        yes Verify battery capacity is  80% of the manufacturer's rating                                              yes capacity < 100% of    (60-month 3.8.6.6          3.8.6.6    when subjected to a performance discharge test or a modified                                      (60-month Frequency manufacturers rating  Frequency performance discharge test                                                                                only)
AND              only) 24 months when battery has reached 85% of the expected life with capacity 100% of manufacturers rating Verify correct inverter voltage, frequency, and alignment to 3.8.7.1          3.8.7.1                                                                            7 days            yes            yes associated 120 VAC buses RS1, RS2, RS3, and RS4.
Verify correct inverter voltage and alignments to required 3.8.8.1          3.8.8.1                                                                            7 days            yes            yes 120 VAC vital buses.
Verify correct breaker alignments to required AC, DC, and 3.8.9.1          3.8.9.1                                                                            7 days            yes            yes 120 VAC bus electrical power distribution subsystems.
Verify correct breaker alignments to required AC, DC, and 3.8.10.1        3.8.10.1                                                                            7 days            yes            yes 120 VAC bus electrical power distribution subsystems.
Section 3.9, Refueling Operations Verify boron concentration is within the limit specified in the 3.9.1.1          3.9.1.1                                                                          72 hours            yes            yes COLR.
3.9.2.1          3.9.2.1    Perform CHANNEL CHECK                                                  12 hours            yes            yes 3.9.2.2          3.9.2.2    Perform CHANNEL CALIBRATION                                          18 months            yes            yes to 1CAN031801 Page 20 of 20 ANO-1      SR Frequency  SR Frequency ANO-1        NUREG-1430                    ANO-1 Surveillance Description Surveillance  Modified by Modified in Proposed SR #          R4 SR #        {NUREG 1430 Description if no ANO-1 Surveillance}
Frequency      TSTF-425  Amendment Request Verify each required reactor building penetration is in the 3.9.3.1        3.9.3.1                                                                          7 days          yes              yes required status.
Verify each required reactor building isolation valve and each 3.9.3.2        3.9.3.2  reactor building purge Isolation valve actuates to the isolation      18 months        yes              yes position Perform CHANNEL CALIBRATION of reactor building purge 3.9.3.3          N/A                                                                          18 months        no              yes exhaust radiation monitor 3.9.4.1        3.9.4.1  Verify one DHR [Decay Heat Removal] loop is in operation              12 hours        yes              yes 3.9.5.1        3.9.5.1  Verify one DHR loop is in operation                                    12 hours        yes              yes Verify correct breaker alignment and indicated power available 3.9.5.2        3.9.5.2                                                                          7 days          yes              yes to each required DHR pump Verify refueling canal water level is  23 feet above the top of 3.9.6.1        3.9.6.1  irradiated fuel assemblies seated within the reactor pressure          24 hours        yes              yes vessel Section 5.5, Programs and Manuals The program shall include the following: Integrated leak test TS 5.5.2.b          N/A                                                                          18 months        no              yes requirements for each system at least once per 18 months.
Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary        18 months on a TSTF-425 TS 5.5.5.d                  during the pressurization mode of operation by one train of the  STAGGERED TEST        yes              yes SR 3.7.10.4 CREVS, operating at the flow rate required by the VFTP, at a            BASIS Frequency of 18 months on a STAGGERED TEST BASIS.
Total particulate concentration of the fuel oil is  10 mg/l when TS 5.5.13.c        N/A    tested every 31 days based on ASTM D 2276, Method A.2 or              31 days          no              yes A.3; N/A 5.5.8          5.5.20                                                                            N/A            yes              yes Surveillance Frequency Control Program}}

Latest revision as of 06:35, 6 January 2025

Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425)
ML18071A319
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 03/12/2018
From: Richard Anderson
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN031801
Download: ML18071A319 (455)
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