ML21362A685

IP 71111.20 Refueling and Other Outage Activities

ML21362A685
Person / Time
Issue date:	12/28/2021
From:	Shawn Campbell NRC/NRR/DRO/IRIB
To:
Stephen.Campbell@nrc.gov
Shared Package
ML21362A677	List: ML21362A685
References
DC 21-026
Download: ML21362A685 (18)

Text

NRC INSPECTION MANUAL IRIB INSPECTION PROCEDURE 71111 ATTACHMENT 20 REFUELING AND OTHER OUTAGE ACTIVITIES Effective Date: 04/06/2022 PROGRAM APPLICABILITY: IMC 2515 A CORNERSTONE: Initiating Events Mitigating Systems Barrier Integrity INSPECTION BASES: See Inspection Manual Chapter (IMC) 0308, Reactor Oversight Process Basis Document, Attachment 2, Technical Basis for Inspection Program SAMPLE REQUIREMENTS:

Sample Requirements Minimum Baseline Sample Budgeted Range Completion Requirements Sample Type Section Frequency Sample Size Samples Hours per Site Refueling Outage Refueling/Other 03.01 One Each Outage* 1 unit: 71+/-15 Outage 2 units: 77+/-15 3 or more units:

81+/-15

• All refueling outages AND outages that enter cold shutdown require use of this Inspection Procedure (IP).

71111.20-01 INSPECTION OBJECTIVES 01.01 To verify that outage activities important to safety are appropriately managed.

01.02 To verify that safety-related and risk-significant structures, systems, and components in inaccessible areas during power operations are appropriately maintained to support required safety functions.

01.03 To verify that risk is appropriately managed during outage-related activities such as during reduced-inventory and midloop conditions.

Issue Date: 04/06/22 1 71111.20

71111.20-02 GENERAL GUIDANCE All refueling outages AND outages that enter cold shutdown require use of this IP. It is at the inspectors judgement and discretion to enter this IP for outages that do not enter cold shutdown and to use applicable elements as determined fit and/or other inspection procedures as noted.

Use Inspection Procedure (IP) 71153, Follow-Up of Events and Notices of Enforcement Discretion, for non-complicated reactor trips when a startup is performed, and the licensee does not enter the containment/drywell. Use IP 71111.19, Post-Maintenance Testing, and IP 71111.22, Surveillance Testing, to address testing activities that normally occur during refueling outages, such as physics testing, emergency diesel generator time response testing, reactor coolant system (RCS) hydrostatic testing, control rod scram time testing, rod drop time testing, reactor trip breaker testing, and containment sump valve testing. Use IP 71111.05, Fire Protection, to perform fire area walkdowns for areas not normally accessible during power operations and to perform a review of PWRs Reactor Coolant Pump oil collection system. Use IP 71111.20 for outage planning and configuration management reviews in areas where it overlaps with other inspection areas, such as maintenance risk assessments and emergent work control, equipment alignment, and inservice inspection activities. Use IP 71111.12, Maintenance Effectiveness as guidance for performing structural walkdowns.

IMC 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process, contains checklists for various plant configurations that can be used to assess mitigating capability. Higher risk configurations have more checklist guidelines for each safety function. The following examples are high-risk configurations:

• pressurized-water reactors (PWRs) breached RCS boundary with steam generators that cannot be used for decay heat removal (DHR) midloop conditions under which it is more likely that DHR can be lost because of poor RCS-level control or poor DHR flow control

• boiling-water reactors (BWRs) cold shutdown conditions (i.e., technical specifications (TS) allow more equipment to be inoperable during cold shutdown conditions than they do during hot shutdown conditions) inoperable safety relief valves under conditions in which such valves would be needed to provide an alternate DHR path and pressure control during a loss of the primary DHR system control rod drive mechanism, or other work on systems attached to the reactor vessel that could result in a rapid loss of reactor vessel water level inventory For each sample, conduct a routine review of problem identification and resolution activities using IP 71152, Problem Identification and Resolution.

Issue Date: 04/06/22 2 71111.20

Sample Considerations Risk Priority Examples Equipment or actions that Inadvertent lowering of the reactor vessel level in midloop could cause a loss of DHR as a result of operator inattention Actions that could adversely Actions that could cause the reactor vessel level indication affect reactor vessel level to be conservative and non-conservative.

Activities that could contribute Activities that affect the electrical power sources to loss of offsite power or designated in the licensees outage/shutdown risk control station blackout plan Equipment used to mitigate a Activities that affect the ability of pumps designated in the loss of DHR licensees outage/shutdown risk control plan as necessary to add water to the reactor vessel Activities that affect the water source for any of the pumps designated in the licensees outage/shutdown risk control plan Equipment used to mitigate a Improper hanging or restoration of clearance tags that loss of reactor vessel level could affect reactor vessel level, DHR, or electrical power availability Failure to verify refueling interlocks Actions that could affect the Exceedance of heatup or cooldown rate limits integrity of the fuel cladding barrier, reactor vessel/RCS, or Failure to establish containment integrity during fuel containment movement 71111.20-03 INSPECTION SAMPLE 03.01 Refueling/Other Outage Sample.

a. Outage Plan.

Before a planned outage, verify that the outage/shutdown risk control plan appropriately considers risk, industry experience, and previous site-specific problems.

Specific Guidance Review the licensees outage/shutdown risk control plan, related industry experience, and previous site-specific problems. Confirm that the licensee has mitigation/response strategies for losses of key safety functions, such as a loss of DHR.

b. Shutdown.

Verify that plant shutdown activities are appropriately conducted. (e.g. verify power reductions are controlled in accordance with licensee procedures)

Issue Date: 04/06/22 3 71111.20

1. Cooldown.

Verify TS cooldown restrictions are followed.

Specific Guidance Observe portions of the cooldown in the control room and review heatup and cooldown curves, calculations, and/or plant process computer data to ensure cooldown is within TS limits.

2. Containment Opening.

If the containment/drywell is opened, and safe for entry (e.g., with radiological controls and industrial safety measures in place) conduct a walkdown of the containment/drywell promptly to assess the condition of systems, structures, and components (SSCs) in the containment/drywell. Review information gathered by videos and photographs and other technologies (such as data collecting-drones, robots) to evaluate inaccessible containment/drywell areas.

At the discretion of the inspectors, the use of this information may be used in lieu of a walkdown.

Specific Guidance If containment entry is available, conduct a thorough walkdown of the containment/drywell as soon as reasonably possible after shutdown to inspect SSCs that are inaccessible during power operations.

(a) Consider the following:

(1) as low as is reasonably achievable controls (2) industrial/personnel safety (heat stress, fire safety awareness)

(3) outage duration (4) unidentified RCS leakage before shutdown (b) During the initial containment walkdown, note areas exhibiting boric acid leakage and verify that the licensee has subsequently identified these areas during the licensees walkdown of containment.

(c) Verify that the containment sump screens are not damaged and are debris free.

(d) Verify that supports, braces, and snubbers are not damaged or deformed. Verify that hydraulic snubbers have no hydraulic fluid leaks and that the hydraulic fluid reservoirs are adequately filled.

(e) Check for items that may be indicative of a larger problem. These problems may include, but are not limited to, the following:

(1) blistered paint and/or rust on containment/drywell liners (2) leakage from containment/drywell ventilation and cooling systems Issue Date: 04/06/22 4 71111.20

(3) cracks in concrete support structures (4) damaged cable insulation (5) presence of loose foreign material or insulation (6) presence of temporary plant equipment left or stored in containment/drywell that the licensees design-basis analyses may not have considered (7) instrument tubing lines that are bent or deformed (8) oil leakage from reactor coolant/recirculation pumps

c. Outage.

Verify that outage-related activities are appropriately conducted.

Specific Guidance Verify that the license maintains defense-in-depth when equipment is taken out of service. Verify that the licensee controls configuration changes in accordance with the outage/shutdown risk control plan and applicable TSs. Consider the effect of concurrent and overlapping work. Confirm that appropriate considerations have been given to scheduling activities and risk mitigating strategies. Such activities may include but are not limited to activities that:

• involve the movement of heavy loads

• could adversely affect the adjacent unit

• can result in a loss of offsite power, compressed air, or cooling water

• involve the erection of scaffolding

• increase the potential for a fire or internal flooding

• impact fuel transfer or the integrity of radiological barriers

1. Isolation and Clearance/Cleanliness Activities.

Verify that tags are properly hung and removed, and that associated equipment is appropriately configured to support the function during clearance restoration.

Specific Guidance Examples of risk-significant clearance activities include the opening of water system boundaries for maintenance near risk-important equipment and power restorations that could adversely affect safety systems, such as the improper repositioning of a motor-operated value upon restoration of power. Pay particular attention to maintenance activities that could affect RCS, DHR, or spent fuel pool cooling.

Issue Date: 04/06/22 5 71111.20

2. Foreign Material Exclusion.

Verify that foreign material exclusion practices are appropriately implemented.

Specific Guidance (a) Review foreign material exclusion logs and records.

(b) Review the controls used to limit and assess debris found inside the containment/drywell/torus/suppression pool. When appropriate, consider the guidance in Operating Experience Smart Sample (OpESS) 2007-01, PWR Containment Sump Recirculation Pipe Foreign Material Blockage. Consider touring these areas to assess cleanliness prior to closeout of the compartment.

3. Reactor Coolant System Instrumentation.

Verify that RCS pressure, level, and temperature instruments have been installed, calibrated, and configured to provide accurate indications.

Specific Guidance For level instruments, verify that the tubing runs do not have elevation changes that could trap either liquid or vapor/gas in the instrument lines (i.e., loop seals). If the plant uses normal operating level instrumentation, verify that the operators have considered the effects of changes in water density (resulting from lower temperature). Verify that the operators are aware of the effect of loss of DHR on the plant's level instrumentation resulting from heatup and pressurization.

For temperature instruments, verify that the operators are aware of the effect of loss of DHR on the plants temperature indication and the potential for discrepancies between the temperature indications and the actual plant state. Temperature may be measured in the DHR loop; in this case interruption, bypass, or partial bypass of DHR flow could lead to inaccurate temperature indications.

4. Electrical Power.

Verify that the status and configurations of electrical systems meet TS requirements and the licensees outage/shutdown risk control plan.

Specific Guidance Verify that switchyard activities are controlled commensurate with safety and are consistent with the licensees outage/shutdown risk control plan assumptions.

5. Containment/Drywell.

Verify that the plant has the proper containment configuration during periods with high decay heat and shorter times to boil and includes provisions for achieving prompt containment closure during periods when containment is permitted to be open.

Issue Date: 04/06/22 6 71111.20

Specific Guidance For BWRs, verify that secondary containment meets the TS requirements. Verify that the containment closure procedures are available and that the licensee uses them when applicable.

For PWRs, verify that containment penetrations meet TS requirements and that containment closure 1 can be achieved at all times.

6. Decay Heat Removal.

Observe DHR parameters to verify proper system operation.

Specific Guidance For BWRs, verify that training and procedures are in place for alternate DHR systems.

For PWRs, when the licensee is relying on the steam generators to provide a backup means of DHR by single-phase natural circulation, verify that the licensee has confirmed the viability of this cooling method.

Some PWRs may allow only one train of residual heat removal (RHR) or shutdown cooling (SDC) to be Operable and in operation in Mode 6 with the water level > 23 feet above the top of the reactor vessel flange. If this condition exists, review Summary of June 24, 2004, Meeting with the Babcock and Wilcox Owners Group and Westinghouse Owners Group to discuss Decay Heat Removal In Pressurized Water Reactors during Refueling with the Upper Reactor Internals Installed (ML041950378) and Westinghouse Owners Group response letter Plant Responses to Address Residual Heat Removal/Shutdown Cooling in Mode 6 with the Reactor Vessel Internals Installed (ML042570147) and verify that the licensee has appropriately managed risk.

When the licensee is relying on the steam generators to provide a backup means of DHR by single-phase natural circulation, verify the following:

(a) Procedures for these methods are derived from analyses, and the required equipment is available.

(b) The RCS pressure boundary is closed.

(c) Steam generator tubes are full.

1 Licensees of PWRs meet containment closure if all containment penetrations (including temporary penetrations, the equipment hatch, and the personnel hatch) have a differential capability equal to the ultimate pressure or would be expected to remain intact following an accident. Leakage requirements, as described in Appendix J, are not a concern. Results from the shutdown probabilistic risk analysis that the Office of Nuclear Regulatory Research conducted at Surry Power Station shows that containment pressure (in a subatmospheric containment) can be high at shutdown after a core damage event.

Issue Date: 04/06/22 7 71111.20

(d) Pressure control capability in the RCS is maintained to ensure the subcooling margin.

(e) The plant has the capability to feed the steam generators.

(f) The plant has the capability to remove steam from the steam generators (e.g., atmospheric relief valves, condenser with steam dump capability).

7. Inventory Control.

Verify that the flowpaths, configurations, and alternative means for inventory addition are consistent with the outage/shutdown risk control plan.

Specific Guidance For activities that have the potential to cause a loss of inventory, verify that adequate controls are in place to prevent inventory loss. RCS pressure boundary problems can pose significant shutdown risk.

(a) Examples of loss of inventory paths include the following:

(1) on BWRsDHR to the suppression pool (2) on BWRsmain steamline paths, including safety relief value removal, automatic depressurization system testing, and main steam isolation valve maintenance (3) on PWRsDHR system crosstie valves, thimble tube seals, and steam generator nozzle dams (4) maintenance activities on connected piping or components that are at an elevation lower than the vessel flange on all plants (5) paths for an intersystem loss-of-coolant accident, such as maintenance and testing on the nonoperating loop low-pressure injection (LPI) train or LPI testing on return back to the refueling water storage tank (b) For BWRs, verify the following:

(1) The automatic isolation on low level is aligned in accordance with TS requirements that apply to the mode.

(2) The main steamline plugs are installed, and the systems required for proper operation of the plugs (e.g., instrument air) are maintained to ensure leak tightness.

(3) The plant has adequate vents to accomplish gravity feed and low-pressure makeup when relied upon.

(4) Examine drain times for various activities that could drain the reactor vessel.

Issue Date: 04/06/22 8 71111.20

8. Reduced Inventory.

Verify that plant conditions during reduced inventory and midloop operations are appropriately controlled.

Specific Guidance Review the unit/outage-specific time-to-boil curves. The time-to-boil period can be less than 30 minutes when DHR is lost under midloop or reduced inventory conditions. During midloop operations, the operator provides the only prevention/mitigating function for a loss of reactor vessel level before the loss of DHR. The plant does not generally have alarms that indicate a loss of level under midloop or reduced inventory conditions. The operators attention to plant conditions is the key preventative aspect for a loss of DHR event. Closely observe the operators performance during drain down, and frequently observe control room activities while the plant is under reduced inventory or midloop conditions. Observe how potential control distractions, such as unexpected conditions and emergent work, affect the operators focus.

For BWR licensees that have adopted reactor pressure vessel water inventory control, review the licensees drain time calculation and calculation assumptions.

Verify that the plant water inventory levels and limiting drain rate are consistent with plant conditions. Verify that TS-required equipment and instrumentation is in the required configuration.

For BWR licensees that have not adopted reactor pressure vessel water inventory control, verify that operations with a potential for draining down the reactor vessel are appropriately managed and comply with TS.

For PWRs, review their commitments from Generic Letter 88-17, Loss of Decay Heat Removal, dated October 17, 1988, and confirm by sampling that they are still in place and adequate. During midloop operations, observe the effect of distractions from unexpected conditions or emergent activities on the operators ability to maintain the required reactor vessel level. In addition to reduced inventory and midloop conditions, assess planned outage activities occurring during other periods when there is a short time-to-boil. Verify the following:

(a) The licensee has reviewed its controls and administrative procedures governing midloop operation and has conducted training for midloop operations.

(b) The licensee uses procedures for the following:

(1) identifying unexpected RCS inventory changes and verifying that an RCS vent path is adequate during and after RCS draining to midloop, IN 88-36 contains information on how a vulnerability exists for a sudden loss of RCS inventory if a plant opens up a cold leg vent without providing an adequate RCS hot leg vent first.

(2) emergency/abnormal operation during reduced inventory (c) Indications of core exit temperature are operable and periodically monitored (typically at least two independent and continuous indications).

Issue Date: 04/06/22 9 71111.20

(d) Indications of RCS water level are operable and periodically monitored (typically at least two independent and continuous indications).

(e) RCS perturbations are avoided.

(f) Means of adding inventory to the RCS are available (typically at least two means in addition to residual heat removal pumps).

(g) Reasonable assurance is obtained that not all hot legs are simultaneously blocked by nozzle dams unless the upper plenum is vented.

(h) Contingency plans exist to repower vital electrical busses from an alternate source if the primary source is lost.

9. Spent Fuel Pool Cooling.

Verify that outage work does not affect the ability of the operations staff to operate the spent fuel pool cooling system during and after core offload consistent with the outage/shutdown risk control plan.

Specific Guidance Verify the following:

(a) Spent fuel pool cooling recovery procedures exist for situations that involve loss of spent fuel pool cooling and are based on current/bounding heat loads.

(b) Operators are trained on backup equipment, power sources, and procedures for loss of spent fuel cooling.

(c) Equipment designated in the recovery procedures is readily available, is dedicated, is unobstructed by outage activities, and has compatible equipment connections.

(d) Instrumentation, alarms, equipment, instructions, and training are provided to alert operators of the need, and to enable them, to add water to the spent fuel pool when necessary.

10. Heavy Lifts.

Verify that heavy load lifts are appropriately conducted.

Specific Guidance See OpESS 2007-03, Crane and Heavy Lift Inspection, Supplemental Guidance to IP 71111.20 and IP 71111.13 for detailed and specific guidance.

11. Refueling.

Verify that reactor core refueling activities are appropriately conducted.

Issue Date: 04/06/22 10 71111.20

Specific Guidance Verify the following:

(a) Fuel handling operations (removal, inspection, sipping, reconstitution, and insertion) and other ongoing activities are being performed in accordance with TS requirements and approved procedures.

(b) Refueling seals have been properly installed and tested, and foreign material exclusion is being maintained in the refueling, spent fuel, and suppression pool areas.

(c) Fuel assembly locations are being tracked, including new fuel, from core offload through core reload. Verify that coupling between the reactivity monitoring instruments and fuel loading location within the core is maintained.

(d) Fuel assembles are loaded in the reactor core locations as specified by the design. Verify fuel assembly loading by reviewing video recording, other core loading records, or physics testing results with as-predicted core operating limit parameters.

(e) Discharged fuel assemblies are placed in allowable locations in the spent fuel pool.

12. Reactivity Control.

Verify that reactivity is being controlled in accordance with the TS. Verify that the outage/shutdown risk control plan identifies activities or structures, systems, and components that could cause unexpected reactivity changes are being appropriately controlled.

Specific Guidance Verify that the licensee has established adequate controls during refueling to preclude improper sequencing of control rods or fuel assemblies to prevent any approach to criticality in any core region without early source range monitor detection.

For PWRs, verify that the licensee has implemented appropriate administrative controls on potential boron dilution paths.

d. Startup.

Verify that plant startup activities are appropriately controlled.

1. Heatup and Startup.

Verify that the licensee has met TS; license conditions; and other requirements, commitments, and administrative procedure prerequisites for mode changes before it changes modes or plant configurations.

Issue Date: 04/06/22 11 71111.20

Specific Guidance (a) Review the reactor physics testing results. Verify that core operating limit parameters are consistent with the design.

(b) Review the RCS boundary leakage calculations. Verify that the licensee has met the TS leakage requirements before making the applicable mode changes.

(c) Review the status of containment/drywell penetrations and containment isolation valves. Verify that the licensee has established containment integrity before entering the applicable TS mode.

(d) Verify that the licensee is appropriately considering and managing plant risk.

2. Containment/Drywell Closure.

If containment/drywell/torus/suppression pool was opened and substantive maintenance activities were performed, then conduct a walkdown or review using other appropriate means, such as photographs, videos, and other technologies (such as data collecting-drones, robots) to evaluate inaccessible containment/drywell areas. If no substantive maintenance activities were performed, and no recent issues related to Items (a) through (e) below, then at the inspectors discretion, no walkdown is needed.

Specific Guidance During walkdowns of those areas, verify the following:

(a) There is no evidence of leakage from pressure boundaries. Focus on SSCs and areas associated with recent maintenance activities.

(b) There is no debris left adrift that might contribute to blockage of the emergency core cooling system strainers.

(c) All items brought into these controlled areas during the outage have either been removed or appropriately evaluated to remain inside.

(d) Maintenance tags have been appropriately cleared.

(e) Passive systems have no obvious indications of damage (e.g., BWR torus suppression pool strainers).

(f) For AP1000 units, verify the proper alignment of the following systems considering the guidance in IP 71111.04, Equipment Alignment:

(1) reactor vessel insulation lower vent (2) Core Makeup Tank (CMT)-A and CMT-B inlet motor operated valve (MOV)

(3) Passive Residual Heat Removal (PRHR) Heat Exchanger (HX) inlet motor operated valve (MOV), and PRHR HX outlet maintenance isolation valve (V109),

(4) In-containment Refueling Water Storage Tank (IRWST) A & B train outlet MOV, Issue Date: 04/06/22 12 71111.20

(5) Accumulator A & B train outlet MOVs (6) IRWST vent covers (7) IRWST gutter drain A & B train air-operated valves (8) Passive Core Cooling System (PXS)-A, PXS-B, and Chemical & Volume Control System (CVS) Compartment Flood Barriers (9) Automatic Depressurization System (ADS) Stage 4 MOVs (10) ADS Stages 1-3 MOVs

e. Fatigue Management.

Verify that the licensee appropriately managed worker fatigue during the outage.

Specific Guidance Verify licensee has complied with Title 10 of the Code of Federal Regulations (10 CFR) 26.205(d)(4) and (5) during the outage. IP 71130.08, Fitness for Duty Program is used to verify work-hour controls for security staff, this review should focus on those staff defined in 10 CFR 26.4(a)(1) through 26.4(a)(4)

Consider the following:

1. Outage Work Schedules. During or after the outage, review actual worked hours by operations, maintenance, and fire brigade personnel during the outage. Verify that (a) Personnel working on outage activities were provided with the minimum days off.

(b) The transition from the operating work schedule into the outage work schedule is managed to ensure all work hour control requirements were met.

(c) Reactor Operators and Senior Reactor Operators on the operating unit, if applicable, are under operating hours control and are not allowed the less restrictive minimum days off requirements for outage activities. Refer to IP 93002, Managing Fatigue, for additional guidance on work hour control requirements.

2. Waiver Requests, Self-Declarations, and Assessments. Once during or after the outage, review three to five waiver requests (10 CFR 26.207, Waivers and Exceptions); self-declarations (10 CFR 26.209, Self-Declarations); and fatigue assessments (10 CFR 26.211, Fatigue Assessments), when they are available.

Verify that the licensee is managing operator fatigue appropriately. IP 93002 provides additional guidance.

71111.20-04 REFERENCES Generic Letter 88-17, Loss of Decay Heat Removal - 10 CFR 50.54(f)

Information Notice 88-36, Possible Sudden Loss of RCS Inventory During Low Coolant Level Operation OpESS 2007-03, Crane and Heavy Lift Inspection, Supplemental Guidance to IP 71111.20 and IP 71111.13 Issue Date: 04/06/22 13 71111.20

OpESS 2007-01, PWR Containment Sump Recirculation Pipe Foreign Material Blockage IMC 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process IP 71111.04, Equipment Alignment IP 71111.05, Fire Protection IP 71111.19, Post-Maintenance Testing IP 71111.22, Surveillance Testing IP 71152, Problem Identification and Resolution IP 71153, Follow-Up of Events and Notices of Enforcement Discretion IP 93002, Managing Fatigue Cross Reference of Generic Communications to IP 71111.20 and Inspection Resources:

http://drupal.nrc.gov/nrr/ope/34016 (nonpublic)

Operating Experience:

http://drupal.nrc.gov/nrr/ope (nonpublic)

IHS Codes and Standards:

https://drupal.nrc.gov/tech-lib/35748 (nonpublic)

NRC Technical Library:

https://drupal.nrc.gov/tech-lib (nonpublic)

END Issue Date: 04/06/22 14 71111.20

Attachment 1 - Revision History for IP 71111.20 Commitment Accession Description of Change Description of Comment Resolution Tracking Number Training and Closed Feedback Number Issue Date Required and Form Accession Change Notice Completion Number Date (Pre-Decisional, Non-Public Information)

N/A 04/03/00 Initial Issuance. Revision history reviewed for the last None CN-00-003 four years N/A ML010820156 Revised to clarify resource estimates for refueling, None 03/06/01 non-refueling, and forced outages. The IP was also CN-01-006 revised to show the interaction between the shutdown operations SDP and IP 71111.13, "Maintenance Risk Assessments and Emergent Work Evaluation," in performing risk assessments in accordance with 10 CFR 50.65(a)(4).

N/A ML020400005 Revised to integrate with IMC 0609, Appendix G, None 01/17/02 "Shutdown Operations Significance Determination CN-02-001 Process," regarding risk insights and how to evaluate inspection findings. It also addresses refueling controls, improper sequencing of control rods or fuel assemblies, reactivity control, and methods to ensure fuel assembles are loaded in correct positions.

N/A ML033230221 Revised to address licensee procedures for foreign None 11/05/03 material exclusion, instruments tracking plant condition CN-03-036 changes, training procedures for BWR alternate decay heat removal. It also cross references IMC 2515, Appendix D, "Plant Status" regarding plant status tours of areas accessible only during outages.

N/A ML041280023 Revised to add guidance for containment inspections, None 05/06/04 when possible, and to add a completion status section.

CN-04-011 Issue Date: 04/06/22 Att 1-1 71111.20

Commitment Accession Description of Change Description of Comment Resolution Tracking Number Training and Closed Feedback Number Issue Date Required and Form Accession Change Notice Completion Number Date (Pre-Decisional, Non-Public Information)

N/A ML052580099 Revised to add the objective of evaluating licensee None 09/02/05 activities during reduced inventory and mid-loop CN-05-024 conditions to ensure that they appropriately manage risk using commitments made in response to GL 88-17 (Loss of Decay Heat Removal). The IP was also revised to recognize that the need for containment walkdown should factor in containment type with regard to ALARA and heat stress.

NA ML080300064 Revised as part of the 2007 ROP realignment to: (1) None ML080250276 01/31/08 clarifies when outages should be addressed via CN-08-005 IP 71111.20 vs IP 71153, and (2) clarifies annual inspection resources for non-refueling and forced outages.

Incorporated guidance regarding: (1) managing fatigue (IP 93002), (2) containment walkthroughs, and (3) inspection resources in accordance with the 2009 ROP realignment. The revised inspection procedure will take effect on January 1, 2010.

NA ML092090694 Guidance regarding (1) managing fatigue (IP 93002), None ML092780024 11/09/09 (2) containment walkthroughs, and (3) inspection CN 09-026 resources in accordance with the 2009 ROP realignment.

Revised to group existing fatigue assessment guidance into a new, separate section, and to add IP 71111.11 as acceptable guidance to observe start-up activities in the control room, based on the 2014 ROP Enhancement Project feedback.

Issue Date: 04/06/22 Att 1-2 71111.20

Commitment Accession Description of Change Description of Comment Resolution Tracking Number Training and Closed Feedback Number Issue Date Required and Form Accession Change Notice Completion Number Date (Pre-Decisional, Non-Public Information)

NA ML15016A138 (1) Grouped existing fatigue assessment guidance into None 71111.20-2046 01/27/15 new, separate section; and (2) added IP 71111.11 as ML15026A551 CN 15-001 acceptable guidance to observe start-up activities in 71111.20-2047 the control room (2014 ROP Enhancement Project ML15026A552 feedback) 71111.20-2048 ML15026A569 ML17179A651 Added guidance to review calculations for reactor None ML17184A001 11/28/17 pressure vessel water inventory control. Eliminated 71111.20-2274 CN 17-027 redundancy and improved plain language. Relocated ML17205A326 optional requirements to the guidance section to better align with the sample completion requirements in IMC 2515, Section 08.04.

ML18130A265 Added heavy load lift review. Revise fatigue None ML18169A319 12/20/18 monitoring section to allow post outage reviews. 71111.20-2041 CN 18-044 Added guidance to assess foreign material controls in ML16161A014 the containment. 71111.20-2042 ML16161A018 71111.20-2149 ML16188A081 71111.20-2188 ML16188A093 ML19179A199 Added AP1000 inspection guidance. Incorporated None ML19317C900 (2019) 10/06/20 additional guidance related to PWR RCS venting and ML20233A521 (2020)

CN 20-047 internal cooling (FBF 71111.20-2402). FBF 71111.20-2402 ML20232C338 ML21013A326 Clarified when the IP should be used, when closeout, None ML21019A075 04/28/21 walkdowns are required and how to perform them, FBF 71111.20-2425 CN 21-020 added scope of review for work-hour control reviews, ML21019A079 and refined AP1000 closeout items. FBF 71111.20-2426 ML21019A080 Issue Date: 04/06/22 Att 1-3 71111.20

Commitment Accession Description of Change Description of Comment Resolution Tracking Number Training and Closed Feedback Number Issue Date Required and Form Accession Change Notice Completion Number Date (Pre-Decisional, Non-Public Information)

N/A ML21362A685 Added guidance to transition to IP 71111.05 during an None ML21362A707 04/06/22 outage, added guidance on verifying drains for CN 22-007 OPDRVs, clarified criteria for containment closure, and FBF 71111.20-2440 provided examples of safe entry conditions into ML21340A185 containment.

Issue Date: 04/06/22 Att 1-4 71111.20