**ATTACHMENT 71111.05T**

INSPECTABLE AREA: Fire Protection (Triennial)

INSPECTION BASES: Fire can be a significant contributor to reactor risk.
In many cases, the risk posed by fires is comparable to or exceeds the
risk from internal events. The fire protection program shall extend the
concept of defense in depth (DID) to fire protection in plant areas
important to safety by:

> \(1\) preventing fires from starting,

(2) rapidly detecting, controlling, and extinguishing fires that do
    occur, and

(3) providing protection for structures, systems, and components
    important to safety so that a fire that is not promptly extinguished
    by fire suppression activities will not prevent the safe shutdown
    (SSD) of the reactor.

LEVEL OF EFFORT: Triennial inspection: Every 3 years, an inspection team
that should include a fire protection inspector, a reactor operations
inspector, and an electrical inspector will conduct a design-based,
plant-specific, risk-informed, onsite inspection of the DID elements
used to mitigate the consequences of a fire. The review will include an
assessment of the licensee=s capability of problem identification and
resolution of fire protection issues.

CHANGES IN SCOPE: For triennial inspections starting March 2006, the
scope of this procedure has been changed to integrate inspection
guidance for operator manual actions associated with licensee-initiated
compensatory measures while the underlying performance deficiency (e.g.
failure to meet the requirements of 10 CFR Part 50, Appendix R, Section
III.G.2 or other plant specific licensing requirements) are corrected.
The background, objectives, and specific guidance are provided in
Section 02.02.11B of this document.

71111.05-01 INSPECTION OBJECTIVES

01.01 The inspection team will evaluate the design, operational status,
and material condition of the licensee=s fire protection program by
verifying that the licensee=s program includes:

-   adequate controls for combustibles and ignition sources inside the
    plant

-   adequate fire detection and suppression capability

-   passive fire protection features in good material condition

-   adequate compensatory measures for out-of-service, degraded or
    inoperable fire protection equipment, systems, or features

-   procedures, equipment, fire barriers, and systems that ensure the
    post-fire capability to safely shut down the plant

-   feasible and reliable manual actions when appropriate to achieve SSD

Inspection Procedure 71111.05AQ, Fire Protection (Annual/Quarterly) is
designed to complement the triennial inspection in the areas of fire
brigade capability and water supply and delivery system maintenance and
adequacy. However, the team should consider the need for additional
inspections in these areas based on previous assessments and potential
issues.

71111.05-02 INSPECTION REQUIREMENTS

02.01 [Inspection Preparation]{.underline} Every 3 years, an inspection
team will select three to five risk-significant fire areas or zones
(depending on the teams makeup, scope, and resources) and conduct
risk-informed inspections of selected aspects of the licensee=s fire
protection program. The team may adjust the number of fire areas
inspected during the inspection based on the complexity of issues.

The initial selection of areas to be inspected should be based on inputs
from a senior reactor analyst (SRA), a fire protection specialist and an
electrical engineer. For each area the selection process will consider
but will not be limited to the following:

\$ a review of the fire hazard analyses

\$ potential ignition sources

\$ the configuration and characteristics of combustible materials

\$ routing of circuits important to accomplish and maintain safe
shutdown condition

\$ the licensee=s fire protection and fire fighting capability

\$ the licensee=s use of operator manual actions

The inspection should focus on post-fire safe shutdown capability and
should inspect alternative or dedicated shutdown capability, as
applicable.

02.02 [Inspection Activities]{.underline} The inspection guidance is
designed to verify that the systems required to achieve and maintain
post-fire SSD are capable of controlling reactivity, reactor coolant
makeup, reactor heat removal, process monitoring, and to support
associated system functions, and that the licensee=s engineering and
licensing documents (e.g., NRC guidance documents, license amendments,
safety evaluation reports (SERs), exemptions, deviations) support the
selection of the designated systems and equipment.

The verification of fixed fire protection systems, including the
installation, design, and testing of the systems, and their adequacy to
control and/or suppress fires in each selected area, will be done
against the code of record.

If a fire brigade drill is observed, the inspection team should consider
the lines of inspection inquiry outlined in IP71111.05QA.

Manual actions not part of an NRC approved exemption or deviation used
in lieu of one of the means specified in Appendix R, Section III.G.2 to
ensure one of the redundant trains is free of fire damage will be
accepted only as temporary compensatory measures and therefore will be
evaluated using guidance provided in paragraph 11B of this section of
the document. If one of the redundant trains in the same fire area is
free of fire damage by one of the specified means in paragraph III.G.2,
then the use of operator manual actions, or other means necessary, to
mitigate fire-induced operation or maloperation to the second train may
be credited without prior approval.

> 1\. [Shutdown From Outside Main Control Room]{.underline}
>
> Verify that hot and cold shutdown can be achieved and maintained from
> outside the control room with or without the availability of offsite
> power for fires in areas where post-fire SSD relies on manipulating
> shutdown equipment from outside the control room.
>
> 2\. [Protection of Safe Shutdown Capabilities]{.underline}
>
> Verify that the fire protection features in place to SSD capability,
> including power, control, and instrumentation cables, satisfy the
> separation and design requirements (for pre-1979 plants Section III.G
> of Appendix R, and for reactor plants reviewed under the Standard
> Review Plan, license specific separation requirements).
>
> 3\. [Passive Fire Protection]{.underline}
>
> Verify through observation of material conditions that the fire
> ratings of fire area boundaries, raceway fire barriers, and equipment
> fire barriers meet the requirements for the fire hazards in the area.
>
> Verify through review of installation or repair records that material
> of an appropriate fire rating (equal to the overall rating of the
> barrier itself) has been used to fill openings and penetrations and
> that the installation meets engineering design.
>
> Verify through review of installation or repair records that material
> of an appropriate fire rating has been used as fire protection wraps
> and that the installation meets engineering design.
>
> For unusual installation configurations and/or application of unusual
> materials verify appropriate fire test data.
>
> 4\. [Active Fire Protection]{.underline}
>
> Verify and review the material condition, operational lineup,
> functionality, and design of fire detection systems, fire suppression
> systems, and manual fire fighting equipment.
>
> Verify that detection, and automatic and manual suppression systems
> are installed, tested, and maintained in accordance with the code of
> record and would adequately control and/or extinguish fires associated
> with the hazards of each selected area.
>
> Verify that the design capability of suppression agent delivery
> systems meet the requirements of the fire hazards.
>
> 5\. [Protection From Damage From Fire Suppression
> Activities]{.underline}
>
> Verify that redundant trains of systems required for hot shutdown,
> which are located in the same fire area, are not subject to damage
> from fire suppression activities or from the rupture or inadvertent
> operation of fire suppression systems, and that the licensee has
> addressed each of the following:
>
> \(A\) A fire in a single location that may, indirectly, through the
> production of smoke, heat, or hot gases, cause activation of automatic
> fire suppression that could potentially damage all redundant trains;
>
> \(B\) A fire in a single location (or inadvertent manual or automatic
> actuation, or rupture of a fire suppression system) that may
> indirectly cause damage to all redundant trains (e.g.,
> sprinkler-caused flooding of other than the locally affected train).
>
> \(C\) Adequate drainage is provided in areas protected by water
> suppression systems.
>
> 6\. [Alternative Shutdown Capability]{.underline}
>
> \(A\) [Methodology]{.underline}
>
> Verify that the licensee\'s alternative shutdown methodology has
> properly identified the systems and components necessary to achieve
> and maintain SSD conditions for each fire area, room or zone selected
> for review. Specifically determine the adequacy of the systems
> selected for reactivity control, reactor coolant makeup, reactor heat
> removal, process monitoring and support system functions.
>
> Verify the licensee=s engineering and/or licensing justifications
> (e.g., NRC guidance documents, license amendments, SERs, exemptions,
> deviations) support the required performance criteria of the selected
> systems and components.
>
> Verify that hot and cold shutdown from outside the control room can be
> achieved and maintained with or without the availability of off-site
> power.
>
> Verify that the transfer of specified plant control functions from the
> control room to the alternative location(s) has been demonstrated
> without being affected by fire‑induced circuit faults (e.g., by the
> use of separate fuses and power supplies for alternative shutdown
> control circuits).
>
> \(B\) [Operational Implementation]{.underline}
>
> Verify that the training program includes an evaluation of alternative
> or dedicated safe shutdown capability for licensed and non‑licensed
> personnel.
>
> Verify that personnel required to place and maintain the plant in hot
> shutdown following a fire using the alternative dedicated shutdown
> system are properly trained and are available at all times from onsite
> shift staff, exclusive of the fire brigade.
>
> Verify that adequate procedures for use of the alternative shutdown
> system are in place.
>
> Verify that human factors attributes were addressed in the development
> of the alternative shutdown procedures (e.g., placement and
> accessability of equipment, environmental conditions, etc.). Consider
> conducting a walk down of the procedure step by step paying special
> attention to the human factors elements.
>
> Verify that the operators can reasonably be expected to perform and
> complete the instructions of the procedures within applicable shutdown
> time requirements. - see Section 11 B
>
> Verify that the licensee conducts periodic operational tests of the
> alternative shutdown transfer capability and instrumentation and
> control functions, and the tests are adequate to prove the
> functionality of the alternative shutdown capability.
>
> 7\. [Circuit Analyses]{.underline}
>
> Verify that the licensee has identified structures, systems, and
> components (SSCs) important to SSD and their demonstrated compliance
> with 10 CFR Part 50.48.
>
> Verify for the selected areas that the licensee has performed a
> post-fire SSD analysis.
>
> Review specific process and instrumentation diagrams (P&IDs) for flow
> diversions, loss of coolant, or other scenarios which could adversely
> affect the nuclear power plants capability to achieve and maintain hot
> shutdown. Verify that the licensee=s analysis identified and
> considered such processes and circuits, and that the analysis has
> shown that hot shorts, and/or shorts to ground will not prevent SSD.
>
> Verify that the circuit analysis considered the following for the
> areas being evaluated:
>
> \(A\) Cable failure modes.
>
> \(i\) For any single thermoplastic or thermoset multiconductor cable
> (including armored), review any combination of conductors within the
> cable (e.g. intra-cable) for which a short will cause spurious
> actuation(s). Inspectors should review 3 or 4 of the most critical
> combinations.
>
> \(ii\) For any 2 adjacent thermoplastic cable, review any combination
> of conductors between the two cables for which a short will cause
> spurious actuation(s). Inspectors should review 3 or 4 of the most
> critical combinations.
>
> \(a\) A maximum of two cables should be evaluated for cases where
> multiple cables may be damaged by the same fire. Multiple spurious
> actuations may be evaluated, depending on the number of conductors,
> and the circuit configuration.
>
> \(b\) For cases involving direct current (DC) control circuits,
> consider the potential spurious operation due to failures of the
> control cables (even if the spurious operation requires two concurrent
> hot shorts of the proper polarity, e.g., plus-to-plus and
> minus-to-minus). Consider potential spurious actuations when the
> source and the target conductors are in two independent multiconductor
> cables.
>
> \(c\) For cases involving decay heat removal (DHR) system isolation
> valves at high-pressure and low-pressure interfaces verify that the
> three-phase power cables to the valves (either thermoset or
> thermoplastic jacketed) are not vulnerable to three-phase proper
> polarity hot shorts.
>
> \(B\) [Common Power Supply/Bus Concern]{.underline}
>
> Verify, on a sample basis, that circuit breaker coordination and fuse
> protection have been analyzed, and are capable of protecting the power
> source of the designated redundant or alternative safe shutdown
> system/equipment.
>
> 8\. [Communications]{.underline}
>
> Verify through inspection of the contents of designated emergency
> storage lockers and review of emergency control station alternative
> shutdown procedures, that portable radio communications and/or fixed
> emergency communications systems are available, operable, and adequate
> for the performance of the designated activities. Assess the
> capability of the communication systems to support the operators in
> the conduct and coordination of their required actions (e.g., consider
> ambient noise levels, clarity of reception, reliability, and coverage
> patterns). If specific, issues arise relating to alternative or
> dedicated shutdown communications adequacy, then, observe a licensee
> conducted communications test in the subject plant area or areas.
>
> Verify that communication equipment such as repeaters, transmitters
> etc. would not be affected by a fire.
>
> 9\. [Emergency Lighting]{.underline}
>
> Review emergency lighting provided, either in fixed or portable form,
> along access routes and egress routes, at control stations, plant
> parameter monitoring locations, and at manual operating stations:
>
> \(A\) If emergency lights are powered from a central battery or
> batteries, verify that the distribution system contains protective
> devices so that a fire in the area will not cause loss of emergency
> lighting in any unaffected area needed for safe shutdown operations.
>
> \(B\) Verify that battery power supplies are rated with at least an
> 8‑hour capacity.
>
> \(C\) Verify the operability testing and maintenance of the lighting
> units follow licensee procedures and accepted industry practice.
>
> \(D\) Verify that sufficient illumination is provided to permit access
> to and verification of components for the monitoring of safe shutdown
> indications and/or the proper operation of SSD equipment.
>
> \(E\) Verify that emergency lighting unit batteries are being
> maintained consistent with the manufacturers recommendations.
>
> 10\. [Cold Shutdown Repairs]{.underline}
>
> Verify that the licensee has repair procedures, equipment, and
> materials to accomplish repairs of components required for cold
> shutdown which might be damaged, that these components can be made
> operable, and that cold shutdown can be achieved within the required
> time frames. Verify that the repair equipment, components, tools, and
> materials (e.g., pre-cut cable connectors with prepared attachment
> lugs) are available and accessible on site.
>
> 11\. [Compensatory Measures.]{.underline}
>
> \(A\) [General guidance]{.underline}. Verify that compensatory
> measures are in place for out-of-service, degraded, or inoperable fire
> protection and post-fire safe shutdown equipment, systems, or features
> (e.g. detection and suppression systems and equipment, passive fire
> barriers, or pumps, valves or electrical devices providing safe
> shutdown functions or capabilities). Short term compensatory measures
> should compensate for the degraded function or feature until
> appropriate corrective action can be taken. Review licensee
> effectiveness in returning the equipment to service in a reasonable
> period of time (typically days or weeks).
>
> If the licensee meets the requirements in 10 CFR Part 50 Appendix R,
> Section III.G.2, then the use of operator manual actions to mitigate
> fire-induced operation or maloperation to the second train may be
> credited without prior approval.
>
> \(B\) [Manual Actions]{.underline}. The three acceptable methods that
> meet the requirement for maintaining one of the redundant trains in
> the same fire area free of fire damage are based on the combination of
> physical barriers, spacial separation, fire detection and automatic
> suppression systems. These methods are described in 10 CFR Part 50
> Appendix R, Section III.G.2. Licensee implemented manual actions to
> respond to potential maloperations that may result from the failure to
> meet this requirement do not correct the underlying performance
> deficiency and therefore will not be accepted as final corrective
> action. However, the staff concluded that for an interim period, while
> appropriate corrective actions are implemented or while preparations
> are made by the licensee to submit exemptions or deviations,
> compensatory measures in the form of manual actions may be acceptable
> if the manual actions meet the criteria provided below.
>
> If the inspectors determine that the manual actions cannot be
> reasonably accomplished and therefore implementation may lead to an
> unsafe plant condition, alternate compensatory measure(s) or temporary
> corrective action(s) must be implemented.
>
> \(i\) Applicability
>
> This guidance is provided for assessing manual actions implemented in
> conjunction with a licensee commitment to Section III.G.2.
>
> Verify that the licensee is committed to meet the requirements of
> Section III.G.2. Determine whether the requirements are met with or
> without the use of manual actions. If manual action are not invoked,
> this guidance is not applicable.
>
> If manual actions were previously approved by the staff and an
> exemption or deviation has been issued, verify that the licensee
> continues to meet the terms of the exemption or deviation.
>
> \(ii\) Diagnostic Instrumentation
>
> Verify that adequate diagnostic instrumentation, unaffected by the
> postulated fire, is provided for the operator to detect the specific
> spurious operation that occurred. Some licensees may have protected
> only the circuits specified in Information Notice 84‑09. Additional
> instrumentation may be needed to properly assess a spurious operation.
> Annunciators, indicating lights, pressure gages, and flow indicators
> are among the instruments typically not protected from the effects of
> a fire. Instrumentation should also be available to verify that the
> manual action accomplished the intended objective.
>
> \(iii\) Environmental Considerations
>
> Evaluate environmental conditions the operators may encounter while
> traveling to the area where the manual action will be performed and
> within the area where the manual action will take place. The
> conditions to be verified may include the following:

-   Radiation levels shall not exceed normal 10 CFR Part 20 limits.

-   Emergency lighting is provided as required in Appendix R, Section
    III.J, or by the licensee=s approved fire protection program.

-   Temperature and humidity conditions are such that they do not affect
    the operators= capability to perform the manual action.

-   Fire effects such as smoke and toxic gases do not affect the
    operators= capability to perform the manual action.

> \(iv\) Staffing
>
> Evaluate licensee shift staffing to determine whether enough qualified
> personnel are available to perform the required manual actions and to
> safely operate the reactor.
>
> \(v\) Communications
>
> Verify that manual action coordination with other plant operations can
> be accomplished, and that communications capability is protected from
> effects of a postulated fire.
>
> \(vi\) Special Tools
>
> Evaluate the need for special tools and verify that such tools are
> dedicated and readily available.
>
> \(vii\) Training
>
> Verify that operator training on the manual actions and the associated
> procedure(s) is adequate and current.
>
> \(viii\) Accessability
>
> Evaluate the accessability of tools and equipment. If special access
> equipment such as ladders are needed, verify the availability of the
> equipment. Verify that an operator can reach the required location
> without personal hazard.
>
> \(ix\) Procedures
>
> Review procedural guidance to ensure that it is adequate and given in
> an emergency procedure. Operators should not rely on having time to
> study normal plant procedures to find a method of operating plant
> equipment that is seldom used.
>
> \(x\) Verification and Validation
>
> Determine whether the manual actions have been verified and validated
> by plant walkdowns using the current procedure. Ensure that the
> licensee has adequately evaluated the capability of operators to
> perform the manual action in the time available before the plant will
> be placed in an unrecoverable condition.

02.03 [Identification and Resolution of Problems]{.underline}. The team
should verify that the licensee is identifying issues related to this
inspection area at an appropriate threshold and entering them in the
corrective action program. For a sample of selected issues documented in
the corrective action program, verify that the corrective actions are
appropriate. See Inspection Procedure 71152, AIdentification and
Resolution of Problems,@ for additional guidance.

71111.05-03 INSPECTION GUIDANCE

03.01 [Inspection Preparation]{.underline}. The team assigned to conduct
the multi-disciplinary triennial fire protection inspection would
include a fire protection inspector, an electrical inspector, and a
reactor operations inspector.

> 1\. [Reactor Operations Inspector (ROI)]{.underline}. The ROI will
> assess the capability of reactor and balance-of-plant systems,
> equipment, operating personnel, and procedures to achieve and maintain
> post-fire safe shutdown and minimize the release of radioactivity to
> the environment in the event of fire. Therefore, the inspection team
> leader will ensure that the ROI is knowledgeable regarding integrated
> plant operations, maintenance, testing, surveillance and quality
> assurance, reactor normal and off-normal operating procedures, and BWR
> and/or PWR nuclear and balance-of-plant systems design.
>
> 2\. [Electrical Inspector (EI)]{.underline}. The EI will identify
> electrical separation requirements for redundant train power, control,
> and instrumentation cables. He will review alternative shutdown panel
> electrical isolation design to establish the panels= electrical
> independence from postulated fire areas. Therefore, the inspection
> team leader will ensure that the EI is knowledgeable regarding reactor
> plant electrical and instrumentation and control (I&C) design and is
> familiar with industry ampacity derating standards.
>
> 3\. [Fire Protection Inspector (FPI)]{.underline}. The FPI will work
> with other team members in determining the effectiveness of the fire
> barriers and systems that establish the reactor plant=s post-fire SSD
> configuration and maintain it free of fire damage. He will determine
> whether suitable fire protection features (suppression, separation
> distance, fire barriers, etc.) are provided for the separation of
> equipment and cables required to ensure plant safety. Therefore, the
> inspection team leader will ensure the FPI is knowledgeable regarding
> reactor plant fire protection systems, features and procedures.

03.02 [Regulatory Requirements and Licensing Bases]{.underline}. The
regulatory requirements and licensing bases against which post-fire safe
shutdown capability is assessed are as follows:

> 1\. [Part 10 of the Code of Federal Regulations]{.underline}. 10 CFR
> 50.48(a), Fire Protection, requires each operating nuclear power plant
> to have a fire protection plan which satisfies the requirements of
> Criterion 3 of Appendix A to 10 Part 50. The NRC has identified that
> an acceptable plan is one that meets the requirements of Appendix R to
> 10 Part 50, or a plan that satisfies the guidance of standard review
> plan (SRP) Section 9.5-1.
>
> 2\. [Plants licensed before January 1, 1979]{.underline}. These plants
> are subject to the requirements of 10 CFR Part 50.48(a) and (b) and
> Appendix R to 10 CFR Part 50. Appendix R, Sections III.G, III.J, and
> III.O were backfit on plants licensed before January 1, 1979.
> Licensees were required to meet the separation requirements of Section
> III.G.2, the alternative or dedicated shutdown capability requirements
> of Sections III.G.3 and III.L, or to request an exemption in
> accordance with 10 CFR Part 50.12. Alternative or dedicated safe
> shutdown capabilities were required to be submitted to the Office of
> Nuclear Reactor Regulation (NRR) for review. NRR approvals are
> documented in SERs.
>
> 3\. [Plants licensed after January 1, 1979]{.underline}. These plants
> are subject to requirements as specified in the conditions of their
> facility operating license, in commitments made to the NRC, or in
> deviations exemptions or licensee amendments granted by the NRC. These
> requirements are generally similar to those in 10 CFR Part 50 Appendix
> R.
>
> 4\. [Changes to the fire protection program]{.underline}. The licensee
> may make changes to the approved fire protection program without prior
> approval by the Commission only if those changes would not adversely
> affect the ability to achieve and maintain safe shutdown in the event
> of a fire. In addition the licensees maybe required to retain the fire
> protection plan and each change to the plan as a record pursuant to
> paragraph 50.48.

03.03 [Inspection Process]{.underline}.

1.  [Licensee Notification Letter]{.underline} The licensee should be
    notified of the triennial inspection in writing at least three
    months in advance of the onsite week. The information gathering
    visit shall be conducted no fewer than three weeks in advance of the
    onsite inspection week. The letter should discuss the scope of the
    inspection, request an information-gathering visit to the licensee
    reactor site/engineering offices, discuss documentation and licensee
    personnel availability needs during the onsite inspection week, and
    request a pre-inspection conference call to discuss administrative
    matters and finalize inspection activity plans and schedules. A
    template for an NRC to licensee triennial fire protection baseline
    inspection notification letter is provided as Attachment 1.

> 2\. [Information-gathering Site Visit]{.underline} The inspection team
> leader should conduct a two to three day information gathering site
> visit. The purposes of the information gathering site visit are to (1)
> gather site-specific information important to inspection planning, and
> (2) conduct initial discussions with licensee representatives
> regarding administrative items and inspection activity plans and
> schedules. In advance of the information-gathering site visit, the
> team leader should provide the licensee with a list of information and
> documents that may be needed for the team to prepare for and conduct
> the triennial inspection, as well as a list of any planned requests
> for licensee conducted evolutions (e.g., emergency lighting tests,
> communication tests, fire drills, shutdown walkthroughs, etc.).
>
> Prior to the inspection information gathering trip, the team leader
> should contact the regional (SRA) to obtain summary of plant specific
> fire risk insights (e.g., fire risk ranking of the rooms/plant fire
> areas, conditional core damage probabilities (CCDPs) for those rooms
> and areas, and transient sequences for these rooms). After considering
> the focus and result of past fire protection and post-fire safe
> shutdown inspections, the team leader should select three to five fire
> areas important to risk for inspection.
>
> After the information gathering site visit, the team leader should use
> the SRA developed fire risk insights, as well as technical input from
> the other team members, to develop an inspection plan addressing (for
> the selected three to five fire areas, zones, as applicable) post-fire
> SSD capability and the fire protection features for maintaining one
> success path of this capability free of fire damage.
>
> 3\. [Information Required/Preparation]{.underline} The team members
> should gather sufficient information to become familiar with the
> following during preparation period:
>
> \(A\) The reactor plant\'s design, layout, and equipment
> configuration.
>
> \(B\) The reactor plant\'s current post-fire safe shutdown licensing
> basis through review of 10 Part 50.48, 10 CFR Part 50 Appendix R (if
> applicable), NRC safety evaluation reports (SERs) on fire protection,
> the plant\'s operating license, updated final safety analysis report
> (UFSAR), and approved exemptions or deviations.
>
> \(C\) The licensee=s strategy and methodology, and derivative
> procedures, for accomplishing post-fire safe shutdown conditions.
> Among the sources of information are the updated final safety analysis
> report (UFSAR), the latest version of the fire hazards analysis (FHA),
> the latest version of the post-fire safe shutdown analysis (SSA), fire
> protection/post-fire safe-shutdown related changes that used 10 Part
> 50.59, 50.48(a) or other criteria, and Generic Letter 86-10 review
> documentation and modification packages, plant drawings,
> emergency/abnormal operating procedures, and the results of licensee
> internal audits (e.g., self assessments and quality assurance (QA)
> audits in the fire protection and post-fire safe shutdown areas).
>
> \(D\) The historical record of plant-specific fire protection issues
> through review of plant-specific documents such as previous NRC
> inspection results, internal audits performed by the reactor licensee
> (e.g., self-assessments and quality assurance audits), corrective
> action system records, event notifications submitted in accordance
> with 10 CFR Part 50.72, and licensee event reports (LERs) submitted in
> accordance with 10 CFR 50.73.
>
> \(E\) The safe shutdown systems and support systems credited by the
> licensee=s analysis for each fire area, room, or zone for
> accomplishing of the required shutdown functions (e.g., reactivity
> control, reactor coolant makeup, reactor heat removal, and process
> monitoring and support functions) as necessary to comply with the SSD
> requirements of 10 CFR Part 50.48(a) and plant-specific licensing
> requirements. The shutdown logic for each area, room, or zone to be
> inspected must be thoroughly understood by the team members.
>
> \(F\) The licensee\'s analytical approach for electrical circuits
> separation analyses, and the licensee=s methodology for identification
> and resolution of circuits of concern. The team=s electrical review
> should include addressing the assumptions and boundary conditions used
> in the performance of the licensee=s analyses.

03.04 [Inspection Activities]{.underline}. For those fire protection
structures, systems, and components installed to satisfy NRC
requirements designed to NFPA codes and standards, the code edition in
force at the time of the design and installation is the code of record
to which the design is evaluated.

Deviations from the codes should be identified and justified in the
UFSAR or FHA. A licensee may apply the equivalency concept in meeting
the provisions of the NFPA codes and standards. When the licensee states
that its design \"meets the NFPA code(s)\" or \"meets the intent of the
NFPA code(s)\" and does not identify any deviations from such codes, the
NRC expects that the design conforms to the codes and the design is
subject to inspection against the NFPA codes. The AAuthority Having
Jurisdiction@ as described in NFPA documents refers to the Director,
Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
Commission, or designee.

If the inspectors determine that the operator manual actions are not
reasonably accomplishable and therefore implementation may not lead to a
safe plant condition, the preliminary finding will be identified and
entered into the Significance Determination Process (SDP), Inspection
Manual Chapter IMC0609 Appendix F.

03.05 [Identification and Resolution of Problems]{.underline}. No
specific guidance is provided.

71111.05-04 RESOURCE ESTIMATE

The resource to perform this inspection procedure is estimated to be 200
hours every 3 years for the triennial inspection regardless of the
number of reactor units at the site.

71111.05-05 PROCEDURE COMPLETION

Inspection of the minimum sample size will constitute completion of this
procedure in the Reactor Programs System (RPS). The minimum sample size
is defined as 3 samples (inspection of three fire areas) regardless of
the number of reactor units at that site:

71111.05-06 INSPECTION RELATED SUPPORTING DOCUMENTATION

This is a broad list of the documents the NRC inspection team may be
interested in reviewing, and possibly obtaining, during the information
gathering site visit.

1\. The current version of the Fire Protection Program and Fire Hazards
Analysis.

2\. Current versions of the fire protection program implementing
procedures (e.g., administrative controls, surveillance testing, fire
brigade).

3\. Fire brigade training program and pre-fire plans.

4\. Post-fire safe shutdown systems and separation analysis.

5\. Post-fire alternative shutdown analysis.

6\. Piping and instrumentation diagrams showing the components used to
achieve and maintain hot standby and cold shutdown for fires outside the
control room and those components used for those areas requiring
alternative shutdown capability.

7\. Plant layout and equipment drawings which identify the physical
plant locations of hot standby and cold shutdown equipment.

8\. Plant layout drawings which identify plant fire area delineation,
areas protected by automatic fire suppression and detection, and the
locations of fire protection equipment.

9\. Plant layout drawings which identify the general location of the
post-fire emergency lighting units.

10\. Plant operating procedures which would be used and describe
shutdown from inside the control room with a postulated fire occurring
in any plant area outside the control room, procedures which would be
used to implement alternative shutdown capability in the event of a fire
in either the control or cable spreading room.

11\. Maintenance and surveillance testing procedures for alternative
shutdown capability and fire barriers, detectors, pumps and suppression
systems.

12\. Maintenance procedures which routinely verify fuse breaker
coordination in accordance with the post-fire safe shutdown coordination
analysis.

13\. A sample of significant fire protection and post-fire safe shutdown
related design change packages and Generic Letter 86-10 evaluations.

14\. The reactor plant\'s IPEEE, results of any post-IPEEE reviews, and
listings of actions taken/plant modifications conducted in response to
IPEEE information.

15\. Temporary modification procedures.

16\. Organization charts of site personnel down to the level of fire
protection staff personnel.

17\. If applicable, layout/arrangement drawings of potential reactor
coolant/recirculation pump lube oil system leakage points and associated
lube oil collection systems.

18\. A listing of the SERs which form the licensing basis for the
reactor plant\'s post-fire safe shutdown configuration.

19\. Procedures/instructions that control the configuration of the
reactor plant\'s fire protection program, features, and post-fire safe
shutdown methodology and system design.

20\. A list of applicable codes and standards related to the design of
plant fire protection features and evaluations of code deviations.

21\. Procedures/instructions that govern the implementation of plant
modifications, maintenance, and special operations, and their impact on
fire protection.

22\. The three most recent fire protection QA audits and/or fire
protection self-assessments.

23\. Recent QA surveillances of fire protection activities.

24\. A listing of open and closed fire protection condition reports
(problem reports/NCRs/EARs/problem identification and resolution
reports).

25\. Listing of plant fire protection licensing basis documents.

26\. A listing of the NFPA code versions committed to (NFPA codes of
record).

27\. A listing of plant deviations from code commitments.

28\. Actual copies of Generic Letter 86-10 evaluations.

71111.05-07 REFERENCES

The SDP Guideline AAppendix 4 - Determining Potential Risk Significance
of Fire Protection and Post-fire Safe Shutdown Inspection Findings.@

Appendix H of the Fire Protection Supplemental Inspection Procedure
(FPSI) AGuidance for Making a Qualitative Assessment of Fire Protection
Inspection Findings, Fire Protection Risk Significance Screening
Methodology@ \[FPRSSM\])

Inspection Procedure 71152, AIdentification and Resolution of Problems.@

Generic Letter 91-18 AInformation to Licensees Regarding Two NRC
Inspection Manual Sections on Resolution of Degraded and Non-conforming
Conditions and on Operability.@

Information Notice 97-48 AInadequate or Inappropriate Interim Fire
Protection Compensatory Measures,@ July 9, 1997

NRC Internal Memorandum dated August 17, 1998, from John N. Hannon to
Arthur T. Howell titled AResponse to Region IV Task Interface Agreement
(TIA) (96TIA008) - Evaluation of Definition of Continuous Fire Watch
(TAC No. M96550).

Individual Plant Examination of Externally Initiated Events(IPEEE)

Regulatory Guide 1.189 AFire Protection for Nuclear Power Plants.@

Regulatory Issue Summary, 2004-03, Rev. 1, ARisk-Informed Approach for
Post-Fire Safe-Shutdown Circuit Inspections.@

Regulatory Issue Summary 2005-07, ACompensatory Measures to Satisfy the
Fire Protection Program Requirements.@

Regulatory Issue Summary 2005-30, AClarification of Post-Fire
Safe-Shutdown Circuit Regulatory Requirements.@

END

ATTACHMENT 1

Mr. President

Licensee Nuclear Department

Licensee Corporation or Company

Address

SUBJECT: SELECTED NUCLEAR POWER STATION, UNITS 1 AND 2 - NOTIFICATION OF
CONDUCT OF A TRIENNIAL FIRE PROTECTION BASELINE INSPECTION

Dear Mr. :

The purpose of this letter is to notify you that the U.S. Nuclear
Regulatory Commission (NRC) Region \# staff will conduct a triennial
fire protection baseline inspection at Selected Nuclear Power Station,
Units 1 and 2 in Month, 20##. The inspection team will be lead by First
Last, a fire protection specialist from the NRC Region \# Office. The
team will be composed of personnel from NRC Region #, and Contracted
National Laboratory. The inspection will be conducted in accordance with
IP 71111.05, the NRC=s baseline fire protection inspection procedure.

The schedule for the inspection is as follows:

! Information gathering visit - Month ##-##, 20## \[Note - this date is
pre-coordinated with the licensee\]

! Week of onsite inspection - Month ##, 20##.

The purposes of the information gathering visit are to obtain
information and documentation needed to support the inspection, to
become familiar with the Selected Nuclear Power Station, Units 1 and 2
fire protection programs, fire protection features, and post-fire safe
shutdown capabilities and plant layout, and, as necessary, obtain plant
specific site access training and badging for unescorted site access. A
list of the types of documents the team may be interested in reviewing,
and possibly obtaining, are listed in Enclosure 1.

During the information gathering visit, the team will also discuss the
following inspection support administrative details: office space size
and location; specific documents requested to be made available to the
team in their office spaces; arrangements for reactor site access
(including radiation protection training, security, safety and fitness
for duty requirements); and the availability of knowledgeable plant
engineering and licensing organization personnel to serve as points of
contact during the inspection.

We request that during the onsite inspection week you ensure that copies
of analyses, evaluations or documentation regarding the implementation
and maintenance of the Selected Nuclear Generating Station, Units 1 and
2 fire protection program, including post-fire safe shutdown capability,
be readily accessible to the team for their review. Of specific interest
are those documents which establish that your fire protection program
satisfies NRC regulatory requirements and conforms to applicable NRC and
industry fire protection guidance. Also, personnel should be available
at the site during the inspection who are knowledgeable regarding those
plant systems required to achieve and maintain safe shutdown conditions
from inside and outside the control room (including the electrical
aspects of the relevant post-fire safe shutdown analyses), reactor plant
fire protection systems and features, and the Selected Nuclear Power
Station fire protection program and its implementation.

Your cooperation and support during this inspection will be appreciated.
If you have questions concerning this inspection, or the inspection
team\'s information or logistical needs, please contact First Last, the
team leader, in the Region \# Office at ###-###-####.

Sincerely,

Docket Nos.: 50-###

and 50-###

Enclosure: As stated (1)

ATTACHMENT 2

Revision History For

Inspection Procedure IP 71111.05T

+-------+-------+---------------------------+-----+-------+------------+
| Commi | Issue | Description of Change     | Tr  | Tra   | Comment    |
| tment | Date  |                           | ain | ining | Resolution |
| Tra   |       |                           | ing | Compl | Accession  |
| cking |       |                           | Nee | etion | Number     |
| N     |       |                           | ded | Date  |            |
| umber |       |                           |     |       |            |
+-------+-------+---------------------------+-----+-------+------------+
| N/A   | 04/   | Previous History Review   | N/A | N/A   | N/A        |
|       | 21/06 |                           |     |       |            |
+-------+-------+---------------------------+-----+-------+------------+
| N/A   | 03/   | Provide inspection        | No  | N/A   | NA         |
|       | 06/03 | guidance to evaluate      |     |       |            |
|       |       | licensee manual actions   |     |       |            |
|       | CN    | which have been           |     |       |            |
|       | 03\   | incorporated into the     |     |       |            |
|       | --007 | procedure as Enclosure    |     |       |            |
+-------+-------+---------------------------+-----+-------+------------+
| N/A   | 12/   | This revised triennial    | Yes | 11/04 | NA         |
|       | 01/04 | fire protection           |     |       |            |
|       |       | inspection procedure      |     |       |            |
|       | CN-04 | includes inspection       |     |       |            |
|       | -027  | guidance for identifying  |     |       |            |
|       |       | circuits that could       |     |       |            |
|       |       | prevent the plant from    |     |       |            |
|       |       | achieving and maintaining |     |       |            |
|       |       | hot shutdown condition    |     |       |            |
|       |       | after a fire. Inspection  |     |       |            |
|       |       | of these circuits was     |     |       |            |
|       |       | suspended in 2000,        |     |       |            |
|       |       | pending the conduct of    |     |       |            |
|       |       | fire tests and the        |     |       |            |
|       |       | assessment of the results |     |       |            |
|       |       | in order to gain risk     |     |       |            |
|       |       | insights into the         |     |       |            |
|       |       | phenomena of fire-induced |     |       |            |
|       |       | electrical cable          |     |       |            |
|       |       | failures. The inspection  |     |       |            |
|       |       | guidance is designed to   |     |       |            |
|       |       | help the inspectors       |     |       |            |
|       |       | identify categories of    |     |       |            |
|       |       | circuit configurations    |     |       |            |
|       |       | most likely to be         |     |       |            |
|       |       | impacted by fire          |     |       |            |
|       |       | potentially affecting the |     |       |            |
|       |       | capability of the         |     |       |            |
|       |       | operators to bring the    |     |       |            |
|       |       | plant to a safe shutdown  |     |       |            |
|       |       | condition.                |     |       |            |
+-------+-------+---------------------------+-----+-------+------------+
| N/A   | 04/   | This revision reflects    | No  |       | NA         |
|       | 21/06 | the withdrawal of the     |     |       |            |
|       |       | Manual Action rule.       |     |       |            |
|       |       | Manual actions will not   |     |       |            |
|       |       | be acceptable as          |     |       |            |
|       |       | alternatives to the       |     |       |            |
|       |       | existing requirements of  |     |       |            |
|       |       | 10 CFR Part 50.48(b)      |     |       |            |
|       |       | unless the licensee       |     |       |            |
|       |       | submits an                |     |       |            |
|       |       | exemption/deviation       |     |       |            |
|       |       | request. However, the use |     |       |            |
|       |       | of manual actions will    |     |       |            |
|       |       | continue to be acceptable |     |       |            |
|       |       | as compensatory measures. |     |       |            |
|       |       | To that effect this       |     |       |            |
|       |       | procedure continues to    |     |       |            |
|       |       | provide guidance to the   |     |       |            |
|       |       | inspectors to assess the  |     |       |            |
|       |       | viability of manual       |     |       |            |
|       |       | actions as compensatory   |     |       |            |
|       |       | measures.                 |     |       |            |
+-------+-------+---------------------------+-----+-------+------------+