UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

Vice President, Operations 

Entergy Nuclear Operations, Inc. 

Palisades Nuclear Plant 

27780 Blue Star Memorial Highway 

Covert, Ml 49043-9530 

August22,2016 

SUBJECT: PALISADES NUCLEAR PLANT- SAFETY EVALUATION REGARDING 

IMPLEMENTATION OF MITIGATING STRATEGIES AND RELIABLE SPENT 

FUEL POOL INSTRUMENTATION RELATED TO ORDERS EA-12-049 AND 

EA-12-051 (CAC NOS. MF0768 AND MF0769} 

Dear Sir or Madam: 

On March 12, 2012, the U.S. Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 

"Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond 

Design-Basis External Events" and Order EA-12-051, "Order to Modify Licenses With Regard 

To Reliable Spent Fuel Pool Instrumentation,'' (Agencywide Documents Access and 

Management System (ADAMS) Accession Nos. ML 12054A 736 and ML 12054A679, 

respectively). The orders require holders of operating reactor licenses and construction permits 

issued under Title 1 O of the Code of Federal Regulations Part 50 to modify the plants to provide 

additional capabilities and defense-in-depth for responding to beyond-design-basis external 

events, and to submit for review Overall Integrated Plans (OIPs) that describe how compliance 

with the requirements of Attachment 2 of each order will be achieved. 

By letter dated February 28, 2013 (ADAMS Accession No. ML 13246A399}, Entergy Nuclear 

Operations, Inc. (Entergy, the licensee) submitted its OIP for Palisades Nuclear Plant 

(Palisades) in response to Order EA-12-049. At six month intervals following the submittal of 

the OIP, the licensee submitted reports on its progress in complying with Order EA-12-049. 

These reports were required by the order, and are listed in the attached safety evaluation. By 

letter dated August 28, 2013 (ADAMS Accession No. ML 13234A503}, the NRG notified all 

licensees and construction permit holders that the staff is conducting audits of their responses 

to Order EA- 12- 049 in accordance with NRC Office of Nuclear Reactor Regulation (NRR) 

Office Instruction LIC-111, "Regulatory Audits" (ADAMS Accession No. ML082900195}. By 

letters dated February 10, 2014 (ADAMS Accession No. ML 13365A264}, and October 13, 2015 

(ADAMS Accession No. ML 15272A324), the NRC issued an Interim Staff Evaluation (ISE) and 

audit report, respectively, on the licensee's progress. By letter dated December 16, 2015 

(ADAMS Accession No. ML 15351 A351 ), Entergy submitted a compliance letter and Final 

Integrated Plan (FIP) in response to Order EA-12-049. The compliance letter stated that the 

licensee had achieved full compliance with Order EA-12-049. 

By letter dated February 28, 2013 (ADAMS Accession No. ML 13060A360), Entergy submitted 

its OIP for Palisades in response to Order EA-12-051. At six month intervals following the 

submittal of the OIP, the licensee submitted reports on its progress in complying with Order 

EA- 12-051. These reports were required by the Order, and are listed in the attached safety 
- 2 -

October 13, 2015 (ADAMS Accession No. ML 15272A324), the NRC staff issued an ISE and 

audit report, respectively, on the licensee's progress. By letter dated March 26, 2014 (ADAMS 

Accession No. ML 14083A620), the NRC notified all licensees and construction permit holders 

that the staff is conducting audits of their responses to Order EA-12-051 in accordance with 

NRC NRA Office Instruction LIC-111, similar to the process used for Order EA-12-049. By letter 

dated December 16, 2015 (ADAMS Accession No. ML 15351A126), Entergy submitted a 

compliance letter in response to Order EA-12-051. The compliance letter stated that the 

licensee had achieved full compliance with Order EA-12-051. 

The enclosed safety evaluation provides the results of the NRC staff's review of Entergy's 

strategies for Palisades. The intent of the safety evaluation is to inform Entergy on whether or 

not its integrated plans, if implemented as described, provide a reasonable path for compliance 

with Orders EA-12-049 and EA-12-051. The staff will evaluate implementation of the plans 

through inspection, using Temporary Instruction 191, "Implementation of Mitigation Strategies 

and Spent Fuel Pool Instrumentation Orders and Emergency Preparedness 

Communications/Staffing/ Multi-Unit Dose Assessment Plans" (ADAMS Accession No. 

ML 15257A188). This inspection will be conducted in accordance with the NRC's inspection 

schedule for the plant. 

If you have any questions, please contact John Hughey, Orders Management Branch, 

Palisades Project Manager, at 301-415-3204 or at John.Hughey@nrc.gov. 

Docket No.: 50-255 

Enclosure: 

Safety Evaluation 

cc w/encl: Distribution via Listserv 

;~ad Mandy K. Halter, Acting Chief 

Orders Management Branch 

Japan Lessons-Learned Division 

Office of Nuclear Reactor Regulation 
TABLE OF CONTENTS 

1~ INTRODUCTION 

2.0 REGULATORY EVALUATION 

2.1 Order EA-12-049 

2.2 Order EA-12-051 

3.0 TECHNICAL EVALUATION OF ORDER EA-12-049 

3.1 Overall Mitigation Strategy 

3.2 Reactor Core Cooling Strategies 

3.2.1 Core Cooling Strategy and PCS Makeup 

3.2.1.1 Core Cooling Strategy 

3.2.1.1.1 Phase 1 

3.2.1.1.2 Phase 2 

3.2.1.1.3 Phase 3 

3.2.1.2 PCS Makeup Strategy 

3.2.1.2.1 Phase 1 

3.2.1.2.2 Phase 2 

3.2.1.2.3 Phase 3 

3.2.2 Variations to Core Cooling Strategy tor Flooding Event 

3.2.3 Staff Evaluations 

3.2.3.1 Availability of Structures, Systems, and Components 

3.2.3.1.1 Plant SSCs 

3.2.3.1.2 Plant Instrumentation 

3.2.3.2 Thermal-Hydraulic Analyses 

3.2.3.3 Reactor Coolant Pump Seals 

3.2.3.4 Shutdown Margin Analyses 

3.2.3.5 FLEX Pumps and Water Supplies 

3.2.3.6 Electrical Analyses 

3.2.4 Conclusions 

3.3 

3.3.1 

3.3.2 

3.3.3 

3.3.4 

3.3.4.1 

3.3.4.1.1 

3.3.4.1.2 

3.3.4.2 

3.3.4.3 

3.3.4.4 

3.3.5 

Spent Fuel Pool Cooling Strategies 

Phase 1 

Phase 2 

Phase 3 

Staff Evaluations 

Availability of Structures, Systems, and Components 

Plant SSCs 

Plant Instrumentation 

Thermal-Hydraulic Analyses 

FLEX Pumps and Water Supplies 

Electrical Analyses 

Conclusions 
3.4 

3.4.1 

3.4.2 

3.4.3 

3.4.4 

3.4.4.1 

3.4.4.1.1 

3.4.4.1.2 

3.4.4.2 

3.4.4.3 

3.4.4.4 

3.4.5 

3.5 

3.5.1 

3.5.2 

3.5.3 

3.5.4 

3.5.5 

3.5.6 

3.6 

3.6.1 

3.6.1.1 

3.6.1.2 

3.6.1.3 

3.6.1.4 

3.6.2 

3.6.3 

3.7 

3.7.1 

3.7.2 

3.7.3 

3.7.3.1 

3.7.3.2 

3.7.4 

3.7.5 

3.7.6 

3.7.7 

3.8 

3.8.1 

3.8.2 

3.8.3 

3.9 

3.9.1 

3.9.1.1 

- 2 -

Containment Function Strategies 

Phase 1 

Phase 2 

Phase 3 

Staff Evaluations 

Availability of Structures, Systems, and Components 

Plant SSCs 

Plant Instrumentation 

Thermal-Hydraulic Analyses 

FLEX Pumps and Water Supplies 

Electrical Analyses 

Conclusions 

Characterization of External Hazards 

Seismic 

Flooding 

High Winds 

Snow, Ice, and Extreme Cold 

Extreme Heat 

Conclusions 

Planned Protection of FLEX Equipment 

Protection from External Hazards 

Seismic 

Flooding 

High Winds 

Snow, Ice, Extreme Cold, and Extreme Heat 

Reliability of FLEX Equipment 

Conclusions 

Planned Deployment of FLEX Equipment 

Means of Deployment 

Deployment Strategies 

FLEX Connection Points 

Mechanical Connection Points 

Electrical Connection Points 

Accessibility and Lighting 

Access to Protected and Vital Areas 

Fueling of FLEX Equipment 

Conclusions 

Considerations in Using Offsite Resources 

Palisades SAFER Plan 

Staging Areas 

Conclusions 

Habitability and Operations 

Equipment Operating Conditions 

Loss of Ventilation and Cooling 
3.9.1.2 

3.9.1.3 

3.9.2 

3.9.2.1 

3.9.2.2 

3.9.2.3 

3.9.3 

3.10 

3.10.1 

3.10.2 

3.10.3 

3.10.4 

3.10.5 

3.11 

3.12 

3.13 

3.14 

3.15 

4.0 

4.1 

4.2 

4.2.1 

4.2.2 

4.2.3 

4.2.4 

4.2.4.1 

4.2.4.2 

4.2.5 

4.2.6 

4.2.7 

4.2.8 

4.2.9 

4.3 

4.3.1 

4.3.2 

4.3.3 

4.4 

- 3 -

Loss of Heating 

Hydrogen Gas Accumulation in Vital Battery Rooms 

Personnel Habitability 

Main Control Room 

Spent Fuel Pool Area 

Other Plant Areas 

Conclusions 

Water Sources 

Steam Generator Make-Up 

Reactor Coolant System Make-Up 

Spent Fuel Pool Make-Up 

Containment Cooling 

Conclusions 

Shutdown and Refueling Analyses 

Procedures and Training 

Maintenance and Testing of FLEX Equipment 

Alternatives to NEI 12-06, Revision O 

Conclusions for Order EA-12-049 

TECHNICAL EVALUATION OF ORDER EA-12-051 

Levels of Required Monitoring 

Evaluation of Design Features 

Design Features: Instruments 

Design Features: Arrangement 

Design Features: Mounting 

Design Features: Qualification 

Augmented Quality Process 

Instrument Channel Reliability 

Design Features: Independence 

Design Features: Power Supplies 

Design Features: Accuracy 

Design Features: Testing 

Design Features: Display 

Evaluation of Programmatic Controls 

Programmatic Controls: Training 

Programmatic Controls: Procedures 

Programmatic Controls: Testing and Calibration 

Conclusions for Order EA-12-051 
- 4 -

5.0 CONCLUSION 

6.0 REFERENCES 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 

RELATED TO ORDERS EA-12-049 AND EA-12-051 

ENTERGY NUCLEAR OPERATIONS, INC. 

PALISADES NUCLEAR PLANT 

DOCKET NO. 50-255 

1.0 INTRODUCTION 

The earthquake and tsunami at the Fukushima Dai-ichi nuclear power plant in March 2011 

highlighted the possibility that extreme natural phenomena could challenge the prevention, 

mitigation and emergency preparedness defense-in-depth layers already in place in nuclear 

power plants in the United States. At Fukushima, limitations in time and unpredictable 

conditions associated with the accident significantly challenged attempts by the responders to 

preclude core damage and containment failure. During the events in Fukushima, the challenges 

faced by the operators were beyond any faced previously at a commercial nuclear reactor and 

beyond the anticipated design-basis of the plants. The U.S. Nuclear Regulatory Commission 

(NRC) determined that additional requirements needed to be imposed at U.S. commercial 

power reactors to mitigate such beyond-design-basis external events (BDBEEs). 

On March 12, 2012, the NRC issued Order EA-12-049, "Order Modifying Licenses with Regard 

to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events" [Reference 

4]. This order directed licensees to develop, implement, and maintain guidance and strategies 

to maintain or restore core cooling, containment, and spent fuel pool (SFP) cooling capabilities 

in the event of a BDBEE. Order EA-12-049 applies to all power reactor licensees and all 

holders of construction permits for power reactors. 

On March 12, 2012, the NRC also issued Order EA-12-051, "Order Modifying Licenses With 

Regard to Reliable Spent Fuel Pool Instrumentation" [Reference 5]. This order directed 

licensees to install reliable SFP level instrumentation with a primary channel and a backup 

channel, and with independent power supplies that are independent of the plant alternating 

current (ac) and direct current (de) power distribution systems. Order EA-12-051 applies to all 

power reactor licensees and all holders of construction permits for power reactors. 

2.0 REGULATORY EVALUATION 

Following the events at the Fukushima Dai-ichi nuclear power plant on March 11, 2011, the 

NRC established a senior-level agency task force referred to as the Near-Term Task Force 

(NTTF). The NTTF was tasked with conducting a systematic and methodical review of the NRC 

regulations and processes and determining if the agency should make additional improvements 

Enclosure 
- 2 -

to these programs in light of the events at Fukushima Dai-ichi. As a result of this review, the 

NTTF developed a comprehensive set of recommendations, documented in SECY-11-0093, 

"Near-Term Report and Recommendations for Agency Actions Following the Events in Japan," 

dated July 12, 2011 [Reference 1 ]. Following interactions with stakeholders, these 

recommendations were enhanced by the NRG staff and presented to the Commission. 

On February 17, 2012, the NRG staff provided SECY-12-0025, "Proposed Orders and Requests 

for Information in Response to Lessons Learned from Japan's March 11, 2011, Great Tohoku 

Earthquake and Tsunami," [Reference 2] to the Commission. This paper included a proposal to 

order licensees to implement enhanced BDBEE mitigation strategies. As directed by the 

Commission in staff requirements memorandum (SRM)-SECY-12-0025 [Reference 3], the NRG 

staff issued Orders EA-12-049 and EA-12-051. 

2.1 Order EA-12-049 

Order EA-12-049, Attachment 2, [Reference 4] requires that operating power reactor licensees 

and construction permit holders use a three-phase approach for mitigating BDBEEs. The initial 

phase requires the use of installed equipment and resources to maintain or restore core cooling, 

containment and SFP cooling capabilities. The transition phase requires providing sufficient, 

portable, onsite equipment and consumables to maintain or restore these functions until they 

can be accomplished with resources brought from off site. The final phase requires obtaining 

sufficient offsite resources to sustain those functions indefinitely. Specific requirements of the 

order are listed below: 

1) Licensees or construction permit (CP) holders shall develop, implement, and 

maintain guidance and strategies to maintain or restore core cooling, 

containment, and SFP cooling capabilities following a beyond-design-basis 

external event. 

2) These strategies must be capable of mitigating a simultaneous loss of all ac 

power and loss of normal access to the ultimate heat sink [UHS] and have 

adequate capacity to address challenges to core cooling, containment, and 

SFP cooling capabilities at all units on a site subject to this Order. 

3) Licensees or CP holders must provide reasonable protection for the 

associated equipment from external events. Such protection must 

demonstrate that there is adequate capacity to address challenges to core 

cooling, containment, and SFP cooling capabilities at all units on a site 

subject to this Order. 

4) Licensees or CP holders must be capable of implementing the strategies in 

all modes of operation. 

5) Full compliance shall include procedures, guidance, training, and acquisition, 

staging, or installing of equipment needed for the strategies. 

On August 21, 2012, following several submittals and discussions in public meetings with NRC 

staff, the Nuclear Energy Institute (NEI) submitted document NEI 12-06, "Diverse and Flexible 

Coping Strategies (FLEX) Implementation Guide," Rev. O [Reference 6] to the NRG to provide 
- 3 -

specifications for an industry-developed methodology for the development, implementation, and 

maintenance of guidance and strategies in response to the Mitigation Strategies order. The 

NRC staff reviewed NEI 12-06 and on August 29, 2012, issued its final version of Japan 

Lessons-Learned Directorate (JLD) Interim Staff Guidance (ISG) JLD- ISG- 2012-01, 

"Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for 

Mitigation Strategies for Beyond-Design-Basis External Events" [Reference 7), endorsing NEI 

12-06, Rev. 0, with comments, as an acceptable means of meeting the requirements of Order 

EA-12-049, and published a notice of its availability in the Federal Register(?? FR 55230). 

2.2 Order EA-12-051 

Order EA-12-051, Attachment 2, [Reference 5) requires that operating power reactor licensees 

and construction permit holders install reliable SFP level instrumentation. Specific requirements 

of the order are listed below: 

All licensees identified in Attachment 1 to the order shall have a reliable 

indication of the water level in associated spent fuel storage pools capable of 

supporting identification of the following pool water level conditions by trained 

personnel: (1) level that is adequate to support operation of the normal fuel pool 

cooling system, (2) level that is adequate to provide substantial radiation 

shielding for a person standing on the spent fuel pool operating deck, and (3) 

level where fuel remains covered and actions to implement make-up water 

addition should no longer be deferred. 

1. The spent fuel pool level instrumentation shall include the following design 

features: 

1.1 Instruments: The instrumentation shall consist of a permanent, fixed 

primary instrument channel and a backup instrument channel. The 

backup instrument channel may be fixed or portable. Portable 

instruments shall have capabilities that enhance the ability of trained 

personnel to monitor spent fuel pool water level under conditions that 

restrict direct personnel access to the pool, such as partial structural 

damage, high radiation levels, or heat and humidity from a boiling pool. 

1.2 Arrangement: The spent fuel pool level instrument channels shall be 

arranged in a manner that provides reasonable protection of the level 

indication function against missiles that may result from damage to the 

structure over the spent fuel pool. This protection may be provided by 

locating the primary instrument channel and fixed portions of the backup 

instrument channel, if applicable, to maintain instrument channel 

separation within the spent fuel pool area, and to utilize inherent shielding 

from missiles provided by existing recesses and corners in the spent fuel 

pool structure. 

1.3 Mounting: Installed instrument channel equipment within the spent fuel 

pool shall be mounted to retain its design configuration during and 

following the maximum seismic ground motion considered in the design of 

the spent fuel pool structure. 
- 4 -

1.4 Qualification: The primary and backup instrument channels shall be 

reliable at temperature, humidity, and radiation levels consistent with the 

spent fuel pool water at saturation conditions for an extended period. 

This reliability shall be established through use of an augmented quality 

assurance process (e.g., a process similar to that applied to the site fire 

protection program). 

1.5 Independence: The primary instrument channel shall be independent of 

the backup instrument channel. 

1.6 Power supplies: Permanently installed instrumentation channels shall 

each be powered by a separate power supply. Permanently installed and 

portable instrumentation channels shall provide for power connections 

from sources independent of the plant ac and de power distribution 

systems, such as portable generators or replaceable batteries. Onsite 

generators used as an alternate power source and replaceable batteries 

used for instrument channel power shall have sufficient capacity to 

maintain the level indication function until offsite resource availability is 

reasonably assured. 

1. 7 Accuracy: The instrument channels shall maintain their designed 

accuracy following a power interruption or change in power source 

without recalibration. 

1.8 Testing: The instrument channel design shall provide for routine testing 

and calibration. 

1.9 Display: Trained personnel shall be able to monitor the spent fuel pool 

water level from the control room, alternate shutdown panel, or other 

appropriate and accessible location. The display shall provide 

on- demand or continuous indication of spent fuel pool water level. 

2. The spent fuel pool instrumentation shall be maintained available and reliable 

through appropriate development and implementation of the following 

programs: 

2.1 Training: Personnel shall be trained in the use and the provision of 

alternate power to the primary and backup instrument channels. 

2.2 Procedures: Procedures shall be established and maintained for the 

testing, calibration, and use of the primary and backup spent fuel pool 

instrument channels. 

2.3 Testing and Calibration: Processes shall be established and maintained 

for scheduling and implementing necessary testing and calibration of the 

primary and backup spent fuel pool level instrument channels to maintain 

the instrument channels at the design accuracy. 
- 5 -

On August 24, 2012, following several NEI submittals and discussions in public meetings with 

NRG staff, the NEI submitted document NEI 12-02, "Industry Guidance for Compliance With 

NRG Order EA-12-051, To Modify Licenses With Regard to Reliable Spent Fuel Pool 

Instrumentation," Rev. 1 [Reference 8] to the NRG to provide specifications for an 

industry-developed methodology for compliance with Order EA-12-051. On August 29, 2012, 

the NRG staff issued its final version of Japan Lessons-Learned Directorate (JLD) Interim Staff 

Guidance (ISG) JLD-ISG-2012-03, "Compliance with Order EA-12-051, Reliable Spent Fuel 

Pool Instrumentation" [Reference 9], endorsing NEI 12-02, Rev. 1, as an acceptable means of 

meeting the requirements of Order EA-12-051 with certain clarifications and exceptions, and 

published a notice of its availability in the Federal Register (77 FR 55232). 

3.0 TECHNICAL EVALUATION OF ORDER EA-12-049 

By letter dated February 28, 2013 [Reference 1 OJ, Entergy Nuclear Operations, Inc. (Entergy, 

the licensee) submitted its Overall Integrated Plan (OIP) for Palisades Nuclear Plant (PNP, 

Palisades) in response to Order EA-12-049. By letters dated August 28, 2013 [Reference 11 ], 

February 28, 2014 [Reference 12], August 28, 2014 [Reference 13], February 27, 2015 

[Reference 14] and August 28, 2015 [Reference 15] (Agencywide Documents Access and 

Management System ADAMS Accession Nos. ML 13241 A234, ML 14059A078, ML 14240A279, 

ML 15062A011 and ML 15240A07 4, respectively), the licensee submitted its first five six-month 

updates to the OIP. By letter dated August 28, 2013 [Reference 16], the NRG notified all 

licensees and construction permit holders that the staff is conducting audits of their responses 

to Order EA-12-049 in accordance with NRG Office of Nuclear Reactor Regulation (NRR) Office 

Instruction LIC-111, "Regulatory Audits" [Reference 23]. By letters dated February 1 O, 2014 

[Reference 17] and October 13, 2015 [Reference 18], the NRG issued an Interim Staff 

Evaluation (ISE) and an audit report on the licensee's progress. By letter dated December 16, 

2015 [Reference 19], the licensee reported that full compliance with the requirements of Order 

EA-12-049 was achieved, and submitted a Final Integrated Plan (FIP). 

3.1 Overall Mitigation Strategy 

Attachment 2 to Order EA-12-049 describes the three-phase approach required for mitigating 

BDBEEs in order to maintain or restore core cooling, containment and SFP cooling capabilities. 

The phases consist of an initial phase (Phase 1) using installed equipment and resources, 

followed by a transition phase (Phase 2) in which portable onsite equipment is placed in service, 

and a final phase (Phase 3) in which offsite resources may be placed in service. The timing of 

when to transition to the next phase is determined by plant-specific analyses. 

While the initiating event is undefined, it is assumed to result in an extended loss of ac power 

(ELAP) with loss of normal access to the LUHS. Thus, the ELAP with the LUHS is used as a 

surrogate for a BDBEE. The initial conditions and assumptions for the analyses are stated in 

NEI 12-06, Section 3.2.1, and include the following: 

1. The reactor is assumed to have safely shut down with all rods inserted (subcritical). 

2. The de power supplied by the plant batteries is initially available, as is the ac power from 

inverters supplied by those batteries; however, over time the batteries may be depleted. 

3. There is no core damage initially. 

4. There is no assumption of any concurrent event. 
- 6 -

5. Because the loss of ac power presupposes random failures of safety-related equipment 

(emergency power sources), there is no requirement to consider further random failures. 

Palisades is a Combustion Engineering (CE) pressurized-water reactor (PWR) with a dry 

ambient-pressure containment. The FIP describes the licensee's three-phase approach to 

mitigate a postulated ELAP event. 

Phase 1 

At the onset of an ELAP, the reactor is assumed to trip from full power. The primary coolant 

pumps coast down and flow in the primary coolant system (PCS) transitions to natural 

circulation. Operators will take prompt actions to minimize PCS inventory losses by isolating 

potential PCS letdown paths. Within 20 minutes of the initiation of the event, primary coolant 

pump (PCP) seal controlled bleedoff (CBO) is isolated to minimize seal leakage. Decay heat is 

removed by steaming from the steam generators (SGs) through the atmospheric dump valves 

(ADVs) or main steam safety valves, and makeup to the SGs is initially provided by the turbinedriven auxiliary feedwater (TDAFW) pump taking suction from the condensate storage tank 

(CST). The water supply for the TDAFW pump is initially from the CST; however with a PCS 

cooldown initiated at 2 hours into the event, this would only provide 4 hours of inventory. 

Therefore, at 1 hour, operators cross-connect the primary system makeup storage tank (T-81) 

with the CST. The tanks are cross-connected so that the T-81 tank inventory can be gravity 

drained into the CST prior to depleting the CST inventory. The combined inventory of the CST 

and T-81 tank will provide a minimum of 8 hours of inventory for makeup to the TDAFW pump 

for PCS decay heat removal, in addition to absorbing the latent heat necessary to support the 

planned PCS cooldown. At approximately 4 hours into the event, operators will begin 

deployment of the portable diesel-driven FLEX pump to the Lake Michigan access area, 

adjacent to the service water intake structure, in preparation for its use to provide SG makeup. 

Once an ELAP has been declared at approximately 1 hour into the event, operators are directed 

to take steps to extend the life of the station batteries by shedding unnecessary loads. 

Completing load shedding by 2 hours conserves station battery life such that a minimum of 8 

hours of functionality can be provided for the ELAP event prior to placing the battery chargers in 

service. 

By 2 hours into the event, operators open the ADVs to commence a cooldown of the PCS. The 

PCS cooldown decreases the differential pressure across the PCP seals to reduce seal leakage 

and prolong seal life, allows for passive injection of borated water from the safety injection tanks 

(SITs), and places the PCS and secondary system in a low pressure state that would allow 

injection of water from low pressure sources available per the licensee's FLEX strategy. During 

the PCS cooldown, core cooling and heat removal will be achieved with the use of the TDAFW 

pump to provide makeup to the SGs for heat removal and ADVs for venting steam from the 

SGs. The plant will cool down at approximately 75 to 100 degrees Fahrenheit/hour (°F/hr) to a 

target temperature no less than 355°F (Tco1d) in the PCS. Control of PCS temperature above 

this limit should be maintained by existing direction in Palisades' emergency procedures to 

terminate depressurization of the SGs at a target pressure of 148 pounds per square inch 

absolute (psia). Steam pressure to the TDAFW pump is controlled by throttling the steam 

supply valve, an air-operated valve (AOV) backed by a nitrogen supply station. 
- 7 -

Control of PCS inventory during Phase 1 is maintained by commencing a plant cooldown, which 

allows passive injection of borated inventory from the nitrogen-pressurized SITs. This strategy 

will ensure that natural circulation is maintained during Phase 1 and therefore PCS makeup via 

FLEX equipment is not required at this stage of the event. 

By 2 hours into the event, operators will begin deployment of the FLEX portable diesel 

generator to restore power to load control centers (LCCs) and other credited equipment. The 

actions for deployment, staging, and connection of the portable FLEX diesel generator will be 

completed by 8 hours into the event to prevent station battery depletion. This allows the station 

batteries to continuously supply power to critical instruments and controls throughout the ELAP 

event. 

Phase 2 

At approximately 4 hours into the event, operators will begin deploying the portable dieseldriven FLEX pump to the Lake Michigan access area (adjacent to the service water intake 

structure) in preparation for its use to provide SG makeup. Flexible hose will be dropped 

through removable grating (at locations currently utilized for the B.5.b pump suction) into the 

intake canal to provide access to the pump suction source, Lake Michigan. Once the pump is 

staged, hoses will be routed to diverse connections on the main feedwater (MFW) system piping 

or the auxiliary feedwater (AFW) system piping prior to using the entire inventory of the CST 

and T-81 tank. 

The licensee stated that one portable diesel-driven FLEX pump can supply the water necessary 

for SG makeup, for spent fuel pool cooling, and for batching borated water in boric acid batching 

tank T-77 (tank T-77). There is ample hose available in either of the FLEX storage buildings to 

go around the turbine building to connect the FLEX pump and the AFW core cooling connection, 

and also to support the hose routings for all three SFP cooling strategies. The planned strategy 

remains to route the hoses through the turbine building, but there is sufficient hose to support 

the alternate routings, for defense-in-depth. 

For PCS inventory control, the transition to the Phase 2 strategy is commenced at 8 hours into 

the event, at which time PNP will have a positive-displacement charging pump re-energized for 

PCS injection. The re-powered charging pump will initially draw borated water from the 

concentrated boric acid storage tanks (BASTs). Prior to depletion of the BASTs, boric acid 

batching operations will begin using boric acid batching tank T-77 (tank T-77). As noted above, 

a supply of water for mixing batches of borated coolant will be provided by the portable dieseldriven FLEX pump. Hose will be routed from the FLEX pump distribution manifold to tank T-77. 

Phase 3 

The operators will continue to implement the strategies from Phase 2, refilling the SGs with the 

portable diesel-driven FLEX pump through either the MFW (primary) or AFW (alternate) 

connections, and removing heat using the ADVs. At 24 hours after event occurrence, off-site 

equipment from the National Strategic Alliance of FLEX Emergency Response (SAFER) 

Response Center (NSRC) will be delivered to the site and is credited with being deployed and 

ready to support the Palisades strategy within 72 hours post-event. The NSRC will provide a 
- 8 -

complete set of FLEX replacement equipment for Palisades' onsite Phase 2 equipment, as well 

as additional equipment to process water, produce borated water, and assist in long-term plant 

recovery (e.g., service water pumps, large generators for energizing all Class 1 E equipment). 

SFP Cooling 

Under non-outage conditions, the licensee stated that the maximum SFP heat load is 4.47 

megawatts thermal (MWt). The licensee calculated that a loss of SFP cooling with this heat 

load from an initial SFP temperature of 140°F results in a time to boil of 5.63 hours. Once 

boiling initiates, the SFP water level will decrease with time as liquid water is vaporized. 

Palisades will plan to provide SFP makeup at 18 hours into the event, prior to the water level 

reaching a height of 15 feet (ft.) above the top of the spent fuel. This is calculated to occur at 

21.3 hours; therefore, completing the equipment line-up for initiating SFP make-up at 18 hours 

is conservative. 

For outage conditions, the worst case SFP heat load is 9.02 MWt. Loss of SFP cooling with this 

heat load from an initial SFP temperature of 140°F results in a time to boil of 3.28 hours. With 

the entire core being located in the SFP, manpower resources otherwise allocated to core 

cooling in non-outage conditions can be allocated to aligning SFP make-up which ensures the 

system alignment can be established prior to the point at which SFP conditions become 

challenged. Assuming the worst case SFP heat load, the SFP water level will boil down to a 

height 15 ft. above the top of the spent fuel in just over 11 hours. The licensee plans to 

complete the equipment line-up necessary for initiating SFP make-up prior to this time. 

The SFP is located in the unit's fuel handling building (FHB) where a vented pathway for steam 

and condensate to exit the FHB is met by opening the double leaf doors on the roof of the FHB. 

The double leaf doors will be opened prior to SFP boiling based on the heat load in the pool at 

the time of the event. The transition to Phase 2 strategies will be made as the inventory in the 

SFP slowly declines due to boiling. The SFP cooling through makeup and spray will be 

provided by using the portable diesel driven FLEX pump with water from Lake Michigan. 

Containment Cooling 

If containment cooling is lost for an extended period of time during an ELAP with the unit in 

Modes 1-4, and no cooling is provided to the containment building, temperature and pressure in 

containment will rise, but remain below design limits for at least 10 days. For Phases 1 and 2 

the licensee's calculations demonstrate no actions are required to maintain containment 

pressure below design limits. 

For Phase 3, the NSRC will supply a low pressure/high flow pump and two 4160 Volts (V)ac 

generators for long-term plant recovery. The low pressure/high flow pump allows for the 

restoration of service water flow to equipment such as the component cooling water (CCW) 

system and low pressure safety injection (LPSI) pump. The 4160 Vac generators allow for the 

restoration of one complete class 1-E distribution bus which provides power to a LPSI pump, a 

CCW pump, shutdown cooling (SOC) flow path valves and three containment air coolers 

(CACs). 
- 9 -

Below are specific details on the licensee's strategies to restore or maintain core cooling, 

containment, and SFP cooling capabilities in the event of a BDBEE, and the results of the staff's 

review of these strategies. The NRC staff evaluated the licensee's strategies against the 

endorsed NEI 12-06, Rev. 0, guidance. 

3.2 Reactor Core Cooling Strategies 

In accordance with Order EA-12-049, licensees are required to maintain or restore cooling to 

the reactor core in the event of an ELAP concurrent with a LUHS. Although the ELAP results in 

an immediate trip of the reactor, sufficient core cooling must be provided to account for fission 

product decay and other sources of residual heat. Consistent with endorsed guidance from 

NEI 12-06, Phase 1 of the licensee's core cooling strategy credits installed equipment (other 

than that presumed lost to the ELAP/LUHS) that is robust in accordance with the guidance in 

NEI 12-06. In Phase 2, robust installed equipment is supplemented by onsite FLEX equipment, 

which is used to cool the core either directly (e.g., pumps and hoses) or indirectly (e.g., FLEX 

electrical generators and cables repowering robust installed equipment). The equipment 

available onsite for Phases 1 and 2 is further supplemented in Phase 3 by equipment 

transported from the NSRCs. 

To adequately cool the reactor core under ELAP conditions, two fundamental physical 

requirements exist: (1) a heat sink is necessary to accept the heat transferred from the reactor 

core to coolant in the PCS and (2) sufficient PCS inventory is necessary to transport heat from 

the reactor core to the heat sink via natural circulation. Furthermore, inasmuch as heat removal 

requirements for the ELAP event consider only residual heat, the PCS inventory should be 

replenished with borated coolant in order to maintain the reactor in a subcritical condition as the 

PCS is cooled and depressurized. 

As reviewed in this section, the licensee's core cooling analysis for the ELAP/LUHS event 

presumes that, per endorsed guidance from NEI 12-06, the reactor would have been operating 

at full power prior to the event. Therefore, the SGs may be credited as the heat sink for core 

cooling during the ELAP/LUHS event. Maintenance of sufficient PCS inventory, despite 

ongoing system leakage expected under ELAP conditions, is accomplished through a 

combination of installed systems and FLEX equipment. The specific means used by the 

licensee to accomplish adequate core cooling during the ELAP/LUHS event are discussed in 

further detail below. The licensee's strategy for ensuring compliance with Order EA-12-049 for 

conditions where the reactor is shut down or being refueled is reviewed separately in Section 

3.11 of this evaluation. 

3.2.1 Core Cooling Strategy and PCS Makeup 

3.2.1.1 Core Cooling Strategy 

3.2.1.1.1 Phase 1 

Following the trips of the reactor and PCPs resulting from the initiating external event, PCS 

temperature and pressure will stabilize at no-load conditions. Core cooling would be 

accomplished by natural circulation flow in the PCS using the SGs as the heat sink. The SG 

inventory makeup would be promptly initiated using the TDAFW pump taking suction from the 

CST, with steam vented via the ADVs or main steam safety valves. 
- 10 -

The fully protected CST will supply sufficient water to the TDAFW pump to cool the reactor core 

for 4 hours using the SGs. Prior to depletion of the CST, operators would align the T-81 tank to 

gravity drain into the CST; the combined water volume of the CST and T-81 tank represents a 

total of 8 hours' worth of inventory for core cooling in the ELAP event. The T-81 tank has been 

modified so as to be fully protected from all applicable external hazards, with the exception of 

large wind-borne missiles (e.g., an automobile). The licensee's FIP states that reasonable 

protection from wind-borne missiles is provided by surrounding structures, the low probability of 

tornadic activity at Palisades, and administrative controls. (The NRC staff's review of this topic 

is further discussed in Sections 3.10.1 and 3.14 of this evaluation.) If the T-81 tank were 

determined to be compromised during an actual event, FLEX Support Guidelines (FSGs) direct 

operators to align alternate suction inventory to the CST from one of several non-robust water 

sources before the CST volume is depleted. 

Following event initiation, operators will verify feedwater flow is aligned to all SGs. Within 2 

hours of event initiation, operators would further take control of the ADVs to commence a 

symmetric PCS cooldown at 75°F/hr within 2 hours of the initiation of the ELAP event. As noted 

in the licensee's FIP, the cooldown would proceed to a target PCS cold leg temperature of no 

less than 355°F. During the audit, the licensee stated that its emergency procedures direct that 

SG depressurization be halted at a pressure of 148 psia. According to the saturation 

relationship for water, this SG pressure should be appropriate to ensure a PCS cold leg 

temperature slightly above 355°F. 

The ADVs at Palisades are air-operated valves. Under ELAP conditions, the supply of gas to 

the valves' pneumatic operators would be interrupted, as both the normal air system and 

backup bulk nitrogen system would be presumed to be unavailable. Therefore, a new backup 

nitrogen station (Nitrogen Station 9) has been installed to permit pneumatic operation of the 

ADVs during an ELAP event. The licensee's FIP states that the new nitrogen station is fully 

protected for all applicable external events. This station is in close proximity to the Main Control 

Room (MCA). Operators will open the ADVs and commence a cooldown of the PCS at 2 hours 

into the event; the nitrogen station has sufficient capacity for approximately 7.1 hours of 

operation after commencement of the cooldown, which covers the 8-hour duration of Phase 1. 

3.2.1.1.2 Phase 2 

The licensee's primary strategy in Phase 2 is to continue to cool the PCS by feeding the SGs 

and releasing steam through the ADVs at a controlled rate to maintain stable PCS conditions. 

Water injection to the SGs will be supplied by a portable, diesel-driven FLEX pump taking 

suction from Lake Michigan. Discharge hoses from the pump will be routed to a hose 

distribution manifold, and then to a primary SG makeup connection on the Main Feedwater 

(MFW) system or an alternate connection on the AFW system. Operators will begin deploying 

this FLEX pump at approximately 4 hours into the event and will have placed it into service by 8 

hours into the event, before the combined inventory of the CST and T-81 tank has been 

depleted. The licensee's FIP states that the FLEX pump will provide a minimum SG makeup 

capacity of 136.5 gallons per minute (gpm) at a SG pressure of 200 psig. In addition to 

providing SG makeup, the pump is sized to support additional aspects of the FLEX strategy, 

including supplying SFP makeup and providing water for mixing borated coolant to inject into 

the PCS. 
- 11 -

In Phase 2, a portable electric motor-driven air compressor is deployed from the FLEX storage 

facility and placed into service to maintain operability of the AOVs before the nitrogen station 

has been depleted. The compressor will be energized and aligned to the AOVs via a 

connection point on the supply piping from Nitrogen Station 9, no later than 8 hours into the 

event. The air compressor would be one of several loads powered by a 400 kilowatt (kW) FLEX 

portable diesel generator. 

3.2.1.1.3 Phase 3 

Although the initial delivery of NSRC equipment is scheduled to reach plant sites within 24 

hours of being requested, the licensee's FIP does not formally credit the availability of NSRC 

equipment until 72 hours after the event begins. The NSRC will supply a large, low-pressure, 

high-flow pump in tandem with a booster pump, which together are capable of providing at least 

5000 gpm from Lake Michigan to the Service Water System (SWS) via a hose connection 

header, effectively restoring access to the UHS. This restoration of SWS flow allows for the use 

of installed SOC heat exchangers tor long-term core cooling and heat removal. In addition, two 

NSRC 4160 Vac generators with a combined capacity of 2 megawatts (MW) will repower one 

complete train of Class 1 E equipment to bring the plant to cold shutdown, including a CCW 

pump and an SOC pump. However, as noted in the licensee's FIP, transitioning to SOC is 

considered a beneficial long-term recovery action, but was not explicitly analyzed as a required 

part of the licensee's strategy tor indefinite coping to satisfy the order. 

Additional equipment supplied by the NSRC to support core cooling includes a mobile water 

treatment unit to improve the quality of the water used tor SG makeup, and a complete set of 

FLEX replacement equipment to provide backup tor the Palisades Phase 2 equipment (e.g., a 

500 gpm SG makeup pump). 

3.2.1.2 PCS Makeup Strategy 

3.2.1.2.1 Phase 1 

Cooldown and depressurization of the PCS significantly extends the expected coping time 

under ELAP/LUHS conditions because it (1) reduces the potential tor damage to PCP seals (as 

discussed in Section 3.2.3.3) and (2) allows coolant stored in the nitrogen-pressurized SITs to 

inject into the PCS to offset system leakage. In Phase 1, operators would depressurize and 

cool down the PCS to a cold leg temperature of 355°F, thereby accomplishing these objectives. 

The licensee's FIP further states that maintaining PCS conditions above this target value would 

prevent SIT nitrogen cover gas injection; therefore, isolation of the SITs is not necessary in 

Phase 1 to ensure that nitrogen intrusion to the PCS is prevented. 

Upon diagnosis of a loss of all ac power, operators will transition to emergency operating 

procedure (EOP)-3.0, "Station Blackout Recovery." Operators will isolate PCS letdown and 

PCP seal CBO flow within 20 minutes of event initiation. These actions greatly reduce PCS 

inventory loss and delay or reduce the temperature rise at the PCP seals following the loss of 

seal cooling. With credit tor passive injection of SIT inventory, the licensee's analyses 

demonstrate that natural circulation in the PCS will be maintained throughout Phase 1 without 

reliance upon FLEX PCS injection. Similarly, Palisades' ELAP analysis shows that the core will 

maintain sufficient Shutdown Margin during Phase 1 without additional boration of the PCS 

using FLEX equipment. 
- 12 -

3.2.1.2.2 Phase 2 

To provide PCS makeup in Phase 2, Palisades plans to use one of its two installed positive 

displacement charging pumps (P-55B or P-55C), each of which has a capacity of 40 gpm. The 

strategy of repowering installed equipment, as opposed to using a portable FLEX pump, is 

identified as an alternative to NEI 12-06 and discussed further in Section 3.14 of this evaluation. 

According to information provided by the licensee, the charging system can be aligned to 

provide diverse flowpaths to the reactor vessel from the discharge of either pump (i.e., the 

normal charging flowpath or the flowpath through high-pressure safety injection (HPSI) Train 2). 

In addition, diverse means of repowering the charging pumps exist: the two pumps are normally 

powered by different safety-related LCCs, LCC-11 and LCC-12; these can be cross-connected 

so that either pump can be powered by the other LCC as an alternate strategy. The LCC itself 

(LCC-11 or LCC-12) would be powered by either LCC-19 or LCC-20, which in turn would be 

connected to one of the two 400 kW portable FLEX diesel generators stored onsite. 

Per the licensee's FIP, the charging pumps will take suction from the BASTs, with sufficient net 

positive suction head provided by gravity. Each BAST contains at least 4500 gallons of highly 

concentrated (greater than 10,000 parts per million (ppm)) boric acid solution. The licensee's 

ELAP analysis shows that with such a highly concentrated borated water source, approximately 

1 .2 hours of boration is sufficient to keep the core sub-critical indefinitely at the target PCS cold 

leg temperature of 355°F with no xenon-135 (which adds negative reactivity) present in the 

core. According to the sequence of events in the licensee's FIP, requisite supporting actions 

would be completed such that borated PCS injection can be commenced via the licensee's 

FLEX strategy no later than 8 hours into the ELAP event. As discussed further in subsequent 

detail, according to the licensee's calculations, initiating primary makeup by this time would 

maintain single-phase natural circulation in the PCS. The high boric acid concentration also 

means that the necessary boration represents a relatively small volume, approximately 2167 

gallons, which minimizes the potential need for PCS letdown. 

The licensee's FIP credits isolation of CBO flow, together with the PCS cooldown and 

depressurization, with minimizing PCS leakage. As noted above, the licensee's Phase 2 

strategy for PCS makeup first draws upon the combined 9000 gallons of highly concentrated 

borated coolant in the BASTs. Once the BASTs are depleted, PCS makeup will be provided via 

boric acid batching using tank T-77, which drains into the BASTs by gravity. In Phase 2, 

untreated water for mixing batches of boric acid solution is supplied from Lake Michigan via the 

portable diesel-driven FLEX pump and hose distribution manifold. A target boron concentration 

of 1720 ppm will be used to ensure adequate boric acid dissolution, even at cold fluid 

temperatures. 

3.2.1.2.3 Phase 3 

The NSRC will provide both a mobile boration unit and a water purification unit, to ensure that 

long-term PCS makeup and boration is accomplished with purified water. The mobile boration 

unit can provide suction to either a Palisades FLEX pump or an NSRC pump to replenish tank 

(T-77) or directly inject to the PCS via connections in either the HPSI or LPSI systems. 

As noted before, the licensee's FIP does not formally credit the use of NSRC-supplied 

equipment until 72 hours into the ELAP event. However, the initial delivery of NSRC equipment 

should arrive at the plant site within 24 hours of notification, with additional equipment deliveries 
- 13 -

following thereafter. As such, the capability to set up NSRC purification and boration equipment 

is expected to exist well before 72 hours into the event. During audit discussions, the licensee 

recognized the desirability of prioritizing the establishment of water purification equipment to 

limit potential impacts of using untreated water from Lake Michigan for PCS. makeup. 

Therefore, the NRG staff expects the licensee to deploy the equipment necessary to provide 

purified water for PCS makeup as soon as permitted by available resources and the overall 

event response prioritization. 

The Palisades FIP indicates that the NSRC will deliver a mobile boration unit to Palisades 

during Phase 3 of the ELAP event. The mobile boration unit can be used to provide PCS 

makeup using the onsite FLEX pumps or NSRC pumps; alternately, the mobile boration unit can 

be used to replenish tank T-77. The availability of the mobile boration unit in Phase 3 helps to 

ensure that sufficient borated makeup can be prepared to maintain PCS inventory. 

In Phase 3, the UHS function of the SWS system would be restored by a high-capacity NSRC 

pump, as discussed in Section 3.2.1.1.3. Restoration of the UHS would allow the plant to 

transfer heat from the reactor core to the SOC heat exchangers in lieu of the SGs. Before the 

PCS can be cooled to SOC entry conditions, however, the SITs must be isolated to prevent 

injection of the nitrogen cover gas into the PCS. The motor control centers (MCCs) associated 

with the SIT isolation valves will be repowered by either the portable FLEX diesel generator or a 

generator supplied by the NSRC. For the purpose of the staff's review, actions associated with 

the transition to SOC are considered recovery actions that are not required to demonstrate 

indefinite coping in response to Order EA-12-049. 

3.2.2 Variations to Core Cooling Strategy for Flooding Event 

In its FIP, Section 2.6.2, the licensee states that Palisades is susceptible to flooding from seiche 

and precipitation hazards. Flooding due to intense precipitation is limited to 6 in. above grade, 

due to overland runoff to Lake Michigan. A seiche (i.e. a standing wave in a lake), would cause 

flooding that would reach a level no higher than 594.1 ft., per the Updated Final Safety Analysis 

Report (UFSAR), and would recede in 30 minutes. On-site FLEX equipment is protected 

against a flood to a level of 594.4, and would be accessed after the seiche has receded. 

Therefore, there are no variations to the core cooling strategy in the event of a flood. Refer to 

Section 3.5.2 of this safety evaluation (SE) for further discussion on flooding. 

3.2.3 Staff Evaluations 

2.3.1 Availability of Structures. Systems, and Components (SSCs) 

Guidance document NEI 12-06 provides guidance that the baseline assumptions have been 

established on the presumption that other than the loss of the ac power sources and normal 

access to the UHS, installed equipment that is designed to be robust with respect to designbasis external events is assumed to be fully available. Installed equipment that is not robust is 

assumed to be unavailable. Below are the baseline assumptions for the availability of SSCs for 

core cooling during an ELAP caused by a BDBEE. 
- 14 -

3.2.3.1.1 Plant SSCs 

Core Cooling - Phase 1 

The licensee's Phase 1 core cooling FLEX strategy relies on the TDAFW pump to remove heat 

from the PCS by providing cooling water flow to both SGs. The Palisades UFSAR Section 

9.7.2.1, Rev. 27, states the TDAFW pump is located in a Design Class 1 portion of the Turbine 

Building. Furthermore, the licensee evaluated the TDAFW pump and support systems using 

PLP-RPT-13-00050, ''Turbine Driven Auxiliary Feedwater (TDAFW) Upgrade and Evaluation for 

FLEX," Rev. 0. The report found the TDAFW system to be robust for all applicable seismic and 

tornado wind and missile hazards with the addition of tornado missile protection for the steam 

exhaust line, stiffening of support walls for Nitrogen Backup Station 2, and additional pipe 

supports in the TDAFW pump room and on the condensate makeup and return line. In addition, 

report PLP-RPT-13-00050, evaluates the flood protection of the TDAFW pump. The staff noted 

that the TDAFW pump room is located below the max expected flood height resulting from a 

seiche but is protected by a water tight door. Nitrogen Backup Stations 1 and 2 provide 

compressed nitrogen for operation of the feedwater control valve and steam inlet valve, which 

were evaluated in PLP-RPT-13-00050 to be robust. However, Palisades Nuclear Plant 

emergency operating procedure EOP Supplement 19, "Alternate Auxiliary Feedwater Methods," 

Rev. 11, directs operators to use local manual operation of the TDAFW pump if control power 

and air are not available. Based on the PNP UFSAR and PLP-RPT-13-00050, the staff finds 

the TDAFW and Nitrogen Backup Stations 1 and 2 are expected to be available at the start of 

an ELAP event consistent with NEI 12-06, Section 3.2.1.3. Further explanation and the NRC 

staff's evaluation of the robustness and availability of water sources for an ELAP event is 

discussed in Section 3.1 O of this SE. 

The licensee's Phase 1 core cooling FLEX strategy relies on the SG ADVs to vent steam from 

the SGs for a controlled cooldown. As described in the Palisades UFSAR, Rev. 29, Table 5.2-3, 

the ADVs are Seismic Category I. Palisades UFSAR, Rev. 26, Section 10.2.1.1, states the two 

atmospheric dump valves are located in each main steam header upstream of the main steam 

isolation valves (MSIVs). Furthermore, UFSAR, Rev. 19, Table 5.2-3 states that the main 

steam system up to and including the MSIVs is Seismic Category I. The FIP states that the 

ADVs are located in the Design Class 1 seismic Auxiliary Building and the staff noted that 

UFSAR, Rev. 29, Section 5.5.1.3.2 states the Auxiliary Building were designed to withstand the 

design-basis tornado missiles. In the Palisades UFSAR, Rev. 21, Section 5.4.1.2, states that 

Design Class 1 structures are flood-protected for the design-basis flood. During the audit 

process, the licensee stated the ADVs and the associated control circuit are powered from the 

Class 1 E batteries, which will not be impacted by the extended load shed performed to extend 

battery capability for approximately 8 hours. The battery chargers will be restored 

approximately 8 hours after the event; therefore ADV controls will not be impacted. 

Following an ELAP event, the pneumatic supply for the ADVs would be supplied from the newly 

installed Nitrogen Backup Station 9 located in the Auxiliary Building, which was determined to 

be seismically robust in EC-47346, "ADV Nitrogen Station," Rev. 0. 

The licensee's Phase 1 core cooling FLEX strategy relies on its CST and tank T-81 as the water 

source for the TDAFW pump. The staff's evaluation of the robustness and availability of the 

CST and tank T-81 for an ELAP event is discussed in SE Section 3.10.1. 
- 15 -

Core Cooling - Phase 2 

The Palisades Phase 2 core cooling strategy continues to use the SGs as the heat sink. 

Palisades will transition to a diesel driven FLEX pump, with water supplied from the UHS, and 

does not rely on any installed plant SSCs other than installed systems with FLEX connection 

points and water sources discussed in SE Sections 3. 7 and 3.10, respectively. 

Core Cooling - Phase 3. 

Palisades' Phase 3 core cooling strategy initially relies on Phase 2 strategies with the NSRC 

equipment providing backup equipment. Once NSRC equipment arrives on site, the licensee's 

Phase 3 core cooling strategy relies on the use of a CCW pump, SOC pump (also the LPSI 

pump) and flow to the SWS system on via a connection to the inlet flange of SWS pump strainer 

F-2C. Palisades UFSAR Table 5.2-3, Rev. 29, indicates that the CCW System, Essential 

Service Water System, and LPSI (SOC) pumps and systems are Design Class 1 components, 

and the FIP also states components in these systems are Design Class 1. The FIP states that 

the CCW and SOC (LPSI) systems are located in the Auxiliary Building and the connection to 

the SWS is located in the Intake Structure. The SWS FLEX connection is discussed in Section 

3.7. 

PCS Inventory Control - Phase 1 

The licensee's Phase 1 RCS inventory control FLEX strategy relies on the PCS system integrity, 

including limited leakage through the PCP seals, and the licensee's analyses demonstrated that 

no FLEX RCS make up is needed prior to 8 hours. 

PCS Inventory Control - Phase 2 

The licensee's Phase 2 PCS inventory control FLEX strategy relies on one of two installed 

positive displacement charging pumps taking suction from the BASTs and discharging into the 

PCS injection or the HPSI lines. The staff considers the repowering of installed equipment for 

PCS inventory control an alternative to NEI 12-06. See Section 3.14 for further discussion of 

the acceptability of this alternative. In its FIP, the licensee states the two charging pumps are 

located in an Auxiliary Building which is robust for all applicable hazards. Furthermore, the 

licensee evaluated the charging pumps, power supplies, cabling and controls using PLP-RPT13-00051, "Boric Acid Storage Tank (BAST) Suction Piping and Component Evaluation for 

FLEX," Rev. O. The report determined the charging pumps, suction piping and controls to be 

robust for all applicable external hazards. As stated in Section 3.1.2.2. of EC-46465, "FLEX 

Basis," Rev. 0, the discharge piping of the charging pumps to the PCS loops and HPSI Train 2 

are located in the Auxiliary Building. Based on the UFSAR, PLP-RPT-13-00050 and EC-46465, 

the staff finds the charging pumps are expected to be available during an ELAP event 

consistent with NEI 12-06, Section 3.2.1.3. The staff's evaluation of the robustness and 

availability of the BASTs for an ELAP event is discussed in SE Section 3.10.2. 

PCS Inventory Control - Phase 3 

The licensee's Phase 3 PCS inventory strategy does not rely on any additional installed plant 

SSCs other than those discussed in Phase 2. 
- 16 -

3.2.3.1.2 Plant Instrumentation 

Per the FIP, the following instrumentation credited for FLEX will be available following the 

stripping of non-essential loads: 

• PCS Hot Leg Temperature 

• PCS Cold Leg Temperature 

• PCS Pressure (Wide Range) 

• PCS Pressure (Narrow Range) 

• Core Exit Thermocouples 

• SIT Level (3 of 4 channels available after load shed) 

• Pressurizer Level (Wide Range) 

• Reactor Vessel Level 

• Neutron Flux (Source Range and Wide Range) 

• Auxiliary Feedwater Flow 

• SG Pressure 

• SG Water Level 

• Containment Pressure 

• Containment Temperature 

• CST Level 

• Battery Capacity/Voltage 

The instrumentation available at Palisades to support the licensee's strategies for core cooling 

and PCS inventory during the ELAP event is consistent with and in some cases exceeds the 

recommendations specified in the endorsed guidance of NEI 12-06. Based upon the 

information provided by the licensee, the NRC staff understands that indication for the above 

instruments would be available and accessible continuously throughout the ELAP event. All of 

these instruments are located within Seismic Category 1 structures and are fully protected 

against all external events. 

As recommended by Section 5.3.3 of NEI 12-06, FSG-7, "Loss of Vital Instrument or Control 

Power," provides instructions and information to obtain readings locally with the use of portable 

FLEX equipment. Furthermore, as described in its FIP, the portable FLEX equipment credited 

in the licensee's mitigating strategies is supplied with the instrumentation necessary to support 

local equipment operation. 

3.2.3.2 Thermal-Hydraulic Analyses 

In the analysis of the ELAP event performed by the Pressurized Water Reactor Owners Group 

(PWROG) in WCAP-17601-P, which the licensee in part relies upon, the Combustion 

Engineering Nuclear Transient (CENTS) code was chosen for the evaluation of CE-designed 

plants such as Palisades. The CENTS code was further used in additional plant-specific 

analysis that was cited in the licensee's FIP as its basis for the FLEX sequence of events 

timeline presented therein. 

The CENTS code, as described in Westinghouse topical report WCAP-15996-A (ADAMS 

Accession No. ML053320174), is a general-purpose thermal-hydraulic computer code that the 

NRC staff has previously reviewed and approved for calculating the behavior of the primary and 
- 17 -

secondary systems of PWRs designed by CE and Westinghouse during non-loss-of-coolant 

accident (LOCA) (non-LOCA) transients. Although CENTS has been approved for performing 

certain design-basis non-LOCA transient analyses, the NRC staff had not previously examined 

its technical adequacy for performing best-estimate simulations of the ELAP event. Therefore, 

in support of mitigating strategy reviews to assess compliance with Order EA-12-049, the NRC 

staff evaluated licensees' thermal-hydraulic analyses, including a limited review of the significant 

assumptions and modeling capabilities of CENTS and other thermal-hydraulic codes used for 

these analyses. 

Based upon this review, the NRC staff questioned whether CENTS and other codes used to 

analyze ELAP scenarios for PWRs would provide reliable coping time predictions in the reflux or 

boiler-condenser cooling phase of the event because of challenges associated with modeling 

complex phenomena that could occur in this phase, including boric acid dilution in the 

intermediate leg loop seals, two-phase leakage through PCP seals, and primary-to-secondary 

heat transfer with two-phase flow in the PCS. In particular, for PWRs with inverted U-tube SGs, 

the reflux cooling mode is said to exist when vapor boiled off from the reactor core flows out the 

saturated, stratified hot leg and condenses on SG tubes, with the majority of the condensate 

subsequently draining back into the reactor vessel in countercurrent fashion. A specific concern 

arose with the use of CENTS for ELAP analysis because the NRC staff's reviews for previous 

non-LOCA applications had imposed a condition limiting the code's heat transfer modeling in 

natural circulation to the single-phase liquid flow regime. This condition was imposed due to the 

lack of benchmarking for the two-phase flow models that would become active in LOCA 

scenarios. Although the PCS leakage rates in an analyzed ELAP event are significantly lower 

than what is typically evaluated for limiting small-break LOCA scenarios, nevertheless, over the 

extended duration of an ELAP event, two-phase natural circulation flow may eventually be 

reached in the PCS, dependent upon the timing of reestablishing PCS makeup. 

In the PWROG's Core Cooling Position Paper, which was provided in a letter dated January 30, 

2013, the PWROG recommended that the reflux or boiler-condenser cooling phase be avoided 

under ELAP conditions because of uncertainties in plant operators' ability to control natural 

circulation following reflux cooling and the impact of dilute pockets of water on criticality. Due to 

the challenge of resolving the above issues within the compliance schedule specified in Order 

EA-12-049, the NRC staff agreed that PWR licensees should provide makeup to the PCS prior 

to entering the reflux or boiler-condenser cooling phase of an ELAP, such that reliance on 

thermal-hydraulic code predictions during this phase of the event would not be necessary. 

However, the PWROG's Core Cooling Position Paper did not fully address the staff's issues 

with CENTS, and in particular, lacked a quantitative definition for the threshold of entry into 

reflux cooling. 

To address the NRC staff's remaining concerns associated with the use of CENTS to simulate 

the two-phase natural circulation flow that may occur during an ELAP for CE-designed PWRs, 

the PWROG submitted a white paper entitled "Westinghouse Response to NRC Generic 

request for additional information (RAI) on CENTS Code in Support of the Pressurized Water 

Reactor Owners Group (PW ROG) (PA-ASC-1187)" (ADAMS Accession No. ML 14218A083). 

This white paper was originally submitted on September 24, 2013, and a revised version was 

resubmitted on November 20, 2013. The white paper focused on comparing several smallbreak LOCA simulations using the CENTS code to analogous calculations performed with the 

CEFLASH-4AS code, which was previously approved for analysis of design-basis small-break 

LOCAs for CE-designed reactors under the conservative Appendix K paradigm. The analyses 
- 18 -

in the CENTS white paper generally showed that CENTS' predictions were similar or 

conservative relative to CEFLASH-4AS for key figures of merit under conditions where natural 

circulation exists in the PCS, including predictions of PCS loop flow rates and the timing of the 

transition to reflux cooling. The NRG staff's review of the analyses in the white paper included 

performing confirmatory analyses with the TRACE code. In particular, the staff's TRACE 

simulations generally showed reasonable agreement with the predictions of CENTS regarding 

the fraction of the initial PCS mass remaining at the transition to reflux cooling. Therefore, as 

documented by letter dated October 7, 2013 (ADAMS Accession No. ML 13276A555), the NRG 

staff endorsed the approach in the PWROG's white paper as an appropriate means for applying 

the CENTS code to beyond-design-basis ELAP analysis, with the limitation that reliance upon 

CENTS is limited to the phase of the event before reflux cooling begins. 

Quantitatively, as proposed in the PWROG's white paper, the threshold for entry into reflux 

cooling is defined as the point at which the 1-hour centered time-average of the flow quality 

passing over the SG tubes' U-bend exceeds one-tenth (0.1) in any PCS loop. Considering this 

criterion relative to the PCS loop flow predictions of both the CENTS and TRACE codes, the 

NRG staff concurred with this definition as a practical expedient, which provides a reasonable 

criterion for determining the threshold of entering reflux cooling for the purpose of analyzing the 

beyond-design-basis ELAP event. Inasmuch as the transition of flow in the PCS loops from 

natural circulation to reflux cooling is a gradual process that typically occurs over multiple 

hours, a quantitatively defined threshold becomes necessary to assure consistent, objective 

treatment. As further discussed in Section 3.2.3.4, a second metric for ensuring adequate 

coping time is associated with maintaining sufficient natural circulation flow in the PCS to 

support adequate mixing of boric acid. 

Applying the one-tenth flow quality criterion to the analyses completed in WCAP-17601-P, the 

November 20, 2013, revision of the PWROG's white paper on CENTS determined ELAP 

coping times prior to entering the reflux cooling mode for each CE reactor included in WCAP17601-P. Unlike the generic calculations performed for reactors designed by other vendors, 

the analysis for CE plants in WCAP-17601-P was generally conducted at a plant-specific level. 

(One exception is that a single analysis was used to represent four plants of similar design, 

including Palisades.) With respect to Palisades, the CENTS white paper reported a coping 

time prior to entering reflux cooling of 24. 7 hours, by which time PCS makeup should be 

provided. Additionally, the licensee performed further plant-specific analysis to support the 

sequence of events in its FIP. In specific, the licensee conservatively determined its time for 

establishing FLEX PCS makeup (i.e., 8 hours) on the basis of ensuring single-phase natural 

circulation flow in the PCS (i.e., prior to the initiation of vapor voids in the SG U-tube bend). 

The licensee's criterion of ensuring single-phase natural circulation in the PCS is conservative 

because two-phase natural circulation flow would be expected to persist in the PCS for a 

significant additional period of time prior to the transition to reflux cooling, and the NRG staff's 

assessment of the CENTS code demonstrated adequate performance in this regime for 

analyzing the beyond-design-basis ELAP event. Finally, setting aside the arbitrary difference 

in applied acceptance criterion (reflux cooling versus single-phase natural circulation), the NRG 

staff's audit found good agreement between the calculated results for Palisades from WCAP17601-P (as reported in the PWROG's CENTS white paper) and those in the licensee's plantspecific calculations. 
- 19 -

The NRC staff further performed confirmatory simulations with the TRACE code for Palisades 

using an input deck generated from a mixture of plant-specific sources and generic information 

applicable to CE reactors. The results of the staff's calculations indicated that approximately 

19 hours should be available prior to the plant entering the reflux cooling mode, thereby 

confirming the appropriateness of the licensee's mitigating strategy. It should further be 

understood that, similar to the simulations for Palisades in both WCAP-17601-P and the 

licensee's plant-specific analysis, the NRC staff's confirmatory calculations are expected to 

significantly underestimate the actual coping time available to Palisades because (1) these 

simulations assumed a PCP seal leakage rate significantly in excess of that subsequently 

endorsed by the NRC staff for the Flowserve N-9000 seals installed at Palisades and (2) no 

credit is taken for CBO isolation. As a result, the NRC staff considers the licensee's strategy 

for ensuring sufficient PCS makeup to avoid reflux cooling as having ample margin for 

mitigating the analyzed ELAP event. 

Therefore, based on the evaluation above, the NRC staff concludes that the licensee's 

analytical approach should appropriately determine the sequence of events for reactor core 

cooling, including time-sensitive operator actions, and evaluate the required equipment to 

mitigate the analyzed ELAP event, including pump sizing and cooling water capacity. 

3.2.3.3 Primary Coolant Pump (PCP) Seals 

Leakage from PCP seals is among the most significant factors in determining the duration that a 

PWR can cope with an ELAP event prior to initiating PCS makeup. An ELAP event would 

interrupt cooling to the PCP seals, resulting in the potential for increased seal leakage and the 

failure of elastomeric a-rings and other components, which could further increase the leakage 

rate. As discussed above, as long as adequate inventory is maintained in the PCS, natural 

circulation can effectively transfer residual heat from the reactor core to the SGs and limit local 

imbalances in boric acid concentration. Along with cooldown-induced shrinkage of the PCS 

inventory, cumulative leakage from PCP seals governs the duration over which natural 

circulation can be maintained in the PCS. Furthermore, the seal leakage rate at the 

depressurized condition can be a controlling factor in determining the flow capacity requirement 

for FLEX pumps to offset ongoing PCS leakage and recover adequate system inventory. 

Flowserve N-9000 4-stage seals are installed on the PCPs at Palisades. The N-9000 is a 

hydrodynamic seal that was developed by Flowserve in the 1980s. One of the design 

objectives for the N-9000 seal was to provide low-leakage performance under loss-of-sealcooling conditions during events such as a station blackout. In support of its customers' efforts 

to address the ELAP event (which similarly involves a loss of seal cooling) in accordance with 

Order EA-12-049, on August 3, 2015, Flowserve submitted to the NRC staff its "White Paper on 

the Response of the N-Seal Reactor Coolant Pump (RCP) Seal Package to Extended Loss of 

All Power (ELAP)," (ADAMS Accession No. ML 15222A366). The N-Seal white paper contains 

information regarding the expected leakage rates over the course of an ELAP event for each 

PWR at which Flowserve N-Seals are currently installed. By letter dated November 12, 2015 

(ADAMS Accession No. ML 1531 OA094), the staff endorsed the leakage rates described in the 

white paper for the beyond-design-basis ELAP event, subject to certain limitations and 

conditions. 
- 20 -

During the audit, the NRG staff requested that the licensee address the status of its 

conformance with the Flowserve white paper and the limitations and conditions in the NRG 

staff's endorsement letter. The licensee responded that the plant design and planned mitigation 

strategy at Palisades are generally consistent with the information assumed in the calculation 

performed by Flowserve, which is summarized in Table 1 of the white paper. However, the 

licensee noted a discrepancy in that the actual cooldown rate during an ELAP event (75°F/hr) 

may be slightly lower than assumed in the white paper (90°F/hr). The NRG staff noted a further 

discrepancy in that the peak cold leg temperature prior to the PCS cooldown assumed in 

Flowserve's analysis (540°F) was slightly less than the saturation temperature corresponding to 

the lowest nominal setpoint for MSSV valve lift pressure (roughly 545°F). However, a 

confirmatory calculation performed by the NRG staff demonstrated that these slight 

discrepancies do not affect the conclusion that Palisades should have adequate seal 

performance with significant margin under ELAP conditions. Based upon its audit review, the 

NRG staff further considered the endorsement letter's condition on the density of the coolant 

leaking from the PCS to be addressed inasmuch as a conservative PCP seal leakage 

assumption was used for the determination of the time to enter reflux cooling. 

According to measured data from Flowserve's 1988 N-Seal station blackout test, following CBO 

isolation at 0.5 hours, over the course of the succeeding period of 6 to 7 hours during which 

CBO isolation was maintained, the average seal leakage rate was slightly less than 0.05 gpm. 

The NRG staff noted that, similar to the test condition, Palisades's mitigating strategy for the 

ELAP event directs early isolation of all CBO flow (i.e., including the CBO relief isolation valve), 

within 20 minutes of event initiation. Although the NRG staff agreed that it is appropriate to 

allow credit for demonstrated performance, during its review of the Flowserve white paper, the 

staff questioned the extrapolation of evidence from a limited test period of 6 to 7 hours to the 

indefinite coping period associated with the ELAP event. Therefore, while the NRG staff 

ultimately agreed with the credit Flowserve's N-Seal white paper allowed for CBO isolation in 

determining the short-term thermal exposure profile of seal elastomers, the staff did not endorse 

direct application of the average leakage rate measured with the CBO isolated in the 1988 test 

for an indefinite period in the absence of demonstrated long-term seal performance. 

Ultimately, the plant-specific calculations performed by Flowserve in its white paper determined 

that the Palisades FLEX scenario does not exceed the design margin demonstrated in the 1988 

station blackout test, such that increased leakage during the ELAP event due to elastomer 

failure or other causes is not expected. Under this condition, according to the values listed in 

Table 3a of the Flowserve white paper, assuming a leakage rate of 1.5 gpm per PCP would be 

appropriate. Considering the plant has 4 PCPs and accounting for another 1 gpm of additional 

PCS leakage would result in a total PCS leakage of 7 gpm. 

Comparing the assumed leakage rates in the licensee's analyses (i.e., including analysis from 

both WCAP-17601-P and its plant-specific calculation) to the endorsed values from the 

Flowserve white paper, the NRG staff observed that the licensee's analyses for determining the 

threshold for entry into reflux cooling did not credit the installation of the Flowserve N-Seals. 

Instead, the leakage rates were based on the assumption that PCP seal leakage would occur at 

an initial rate of 15 gpm (i.e., a rate just below that which would trigger closure of excess flow 

check valves in the CBO lines) at the PCS temperature and pressure conditions applicable 

when PCP inlet subcooling decreases below 50°F. Modeling PCP seal leakage in this manner 

was intended to envelop the potential for seal instability at low inlet subcooling conditions to 

result in "pop-open" seal failures. Thermal-hydraulic analysis indicates that the subcooling 
- 21 -

margin decreases below S0°F at approximately 3 hours into the event. At this juncture, the PCS 

cooldown is being conducted, which results in the PCS approaching saturation because the 

pressurizer heaters are presumed to lose power during the ELAP event. As the PCS cooldown 

and depressurization continue, the seal leakage rate is assumed to decrease in accordance 

with the choked flow correlation used in the licensee's CENTS analyses. Comparing the 

leakage rates from the licensee's CENTS analyses, as well as confirmatory analysis performed 

by the NRC staff with the TRACE code (see Section 3.2.3.2 above), the NRC staff concluded 

that the analytically assumed leakage rates are conservative relative to the expected leakage 

rate for Flowserve N-9000 seals during the analyzed ELAP event. 

The NRC staff further reviewed the potential PCS leakage rate of 7 gpm relative to the makeup 

rate the licensee could supply using the T-77 batching tank. The licensee's FIP submittal 

asserts that batching a volume of 457 gallons every 90 minutes would be sufficient to account 

for PCS leakage. However, the NRC staff concluded that the rate of injection implied by the 

licensee's statement (i.e., approximately 5 gpm) may not be sufficient to match the long-term 

leakage rates endorsed in the Flowserve N-Seal white paper. As a result, the NRC staff 

requested during the audit that the licensee identify the actual amount of time required to 

complete one batch addition from the T-77 tank. In response, the licensee estimated that one 

batch addition of PCS makeup could be completed via this method in less than 30 minutes. 

This information provides confidence that the licensee should have adequate onsite FLEX PCS 

makeup capability to mitigate potential system leakage rates considered bounding in the 

Flowserve N-Seal white paper. 

Based upon the discussion above, the NRC staff concludes that the PCP seal leakage rates 

assumed in the licensee's thermal-hydraulic analysis may be applied to the beyond-designbasis ELAP event for the site. 

3.2.3.4 Shutdown Margin Analyses 

In an analyzed ELAP event, the loss of electrical power to control rod drive mechanisms is 

assumed to result in an immediate reactor trip with the full insertion of all control rods into the 

core. The insertion of the control rods provides sufficient negative reactivity to achieve 

subcriticality at post-trip conditions. However, as the ELAP event progresses, the shutdown 

margin for PWRs is typically affected by several primary factors: 

• the cooldown of the RCS and fuel rods adds positive reactivity 

• the concentration of xenon-135, which (according to the core operating history 

assumed in NEI 12-06) would 

o initially increase above its equilibrium value following reactor trip, thereby 

adding negative reactivity 

o peak at roughly 12 hours post-trip and subsequently decay away gradually, 

thereby adding positive reactivity 

• the passive injection of borated makeup from nitrogen-pressurized accumulators due 

to the depressurization of the RCS, which adds negative reactivity 
- 22 -

At some point following the cooldown of the PCS, PWR licensees' mitigating strategies 

generally require active injection of borated coolant via FLEX equipment. In many cases, 

boration would become necessary to offset the gradual positive reactivity addition associated 

with the decay of xenon-135; but, in any event, borated makeup would eventually be required to 

offset ongoing PCS leakage. The necessary timing and volume of borated makeup depend on 

the particular magnitudes of the above factors for individual reactors. 

In support of its review of the mitigating strategy for Palisades, the NRC staff audited the 

licensee's shutdown margin calculation. The licensee's calculation considered a case where 

borated coolant at a concentration of 1720 ppm boron is injected at flow rate of 30 gpm starting 

at 13 hours into the event. The licensee concluded that maintaining this PCS makeup flow for a 

duration of 7 hours will be sufficient to maintain adequate reactor shutdown margin indefinitely 

at the target PCS cold leg temperature of 355°F with no xenon remaining in the core. Based 

upon this calculation, Palisades will conservatively begin PCS boration at 8 hours into the event, 

and will initially draw suction from the BASTs, which contain coolant with a boron concentration 

of at least 10,000 ppm. As described in the sequence of events in the licensee's FIP, the 

licensee's FLEX strategy is capable of providing indefinite coping capability at the conditions 

above, without cooling the RCS to cold shutdown mode. 

As described in its FIP, the licensee's shutdown margin calculation for Palisades assumed 

uniform mixing of boric acid. To support its plant-specific audit reviews, the NRC staff had 

requested that the industry provide additional information to justify that borated makeup would 

adequately mix with the PCS volume under natural circulation conditions potentially involving 

two-phase flow. In response, the PWROG submitted a position paper, dated August 15, 2013 

(withheld from public disclosure due to proprietary content), which provided test data regarding 

boric acid mixing under single-phase natural circulation conditions and outlined applicability 

conditions intended to ensure that boric acid addition and mixing during an ELAP would occur 

under conditions similar to those for which boric acid mixing data is available. By letter dated 

January 8, 2014 (ADAMS Accession No. ML 13276A 183), the NRC staff endorsed the above 

position paper with three conditions: 

• The required timing and quantity of borated makeup should consider 

conditions with no RCS leakage and with the highest applicable leakage rate. 

• Adequate borated makeup should be provided either (1) prior to the RCS 

natural circulation flow decreasing below the flow rate corresponding to 

single-phase natural circulation, or (2) if provided later, then the negative 

reactivity from the injected boric acid should not be credited until one hour 

after the flow rate in the RCS has been restored and maintained above the 

flow rate corresponding to single-phase natural circulation. 

• A delay period adequate to allow the injected boric acid solution to mix with 

the RCS inventory should be accounted for when determining the required 

timing for borated makeup. Provided that the flow in all loops is greater than 

or equal to the corresponding single-phase natural circulation flow rate, a 

mixing delay period of one hour is considered appropriate. 
- 23 -

The NRG staff's audit review considered whether the licensee had followed recommendations 

from the PWROG's position paper and the associated conditions imposed in the NRG staff's 

endorsement letter. In particular, the NRG staff concluded that the licensee had considered a 

bounding range of PCS leakage rates, including a case with no PCS leakage. The NRG staff 

further concluded that initiating PCS makeup at a rate of 30 gpm by 8 hours into the event 

should ensure that adequate natural circulation flow exists in the PCS throughout the period that 

boric acid mixing would be credited. Finally, based on our audit review, the staff agreed that 

initiating boration at 8 hours would provide margin to the timeline defined in the licensee's 

shutdown margin calculation, even after accounting for a 1-hour delay period to ensure 

adequate mixing. Therefore, the NRG staff concluded that the licensee's boration strategy 

conforms to the PWROG position paper, including the additional conditions imposed in the NRG 

staff's endorsement letter. 

The licensee's shutdown margin calculations also included analysis to determine minimum 

quantities of SIT injection in order to satisfy shutdown margin requirements, as well as 

maximum quantities of SIT injection for determining the PCS pressure necessary to prevent 

injection of the nitrogen cover gas into the PCS. In particular, the licensee's calculations 

determined that the SIT isolation valves should be closed prior to depressurizing the PCS 

pressure below approximately 144 psia in Phase 3, which is the approximate saturation 

pressure corresponding to the planned post-PCS-cooldown cold leg temperature of 355°F. 

With regard to assessing the potential need for PCS venting to support injection of the volume 

of borated coolant necessary to ensure adequate shutdown margin, the licensee's FIP states 

that the 10,000 ppm boron concentration of the initial PCS injection source (i.e. the BASTs) 

results in only having to inject approximately 2167 gallons of coolant, which can be 

accommodated by the PCS without letdown, given the nominal pressurizer steam volume of 700 

cubic ft. (over 5000 gallons). The NRG staff further observed that, inasmuch as the FLEX 

strategy directs that the PCS be cooled and depressurized prior to initiating PCS makeup, 

cooldown-induced contraction of the PCS inventory would create significant additional free 

volume beyond that existing during normal operation. Nevertheless, the NRG staff did not 

consider this justification adequate because the injection of coolant from the SITs could result in 

considerable refilling of the PCS in ELAP scenarios considering minimal PCS leakage. Thus, 

the potential for undesirable increases in PCS pressure at reduced PCS temperatures could not 

be ruled out in certain scenarios (e.g., a high-pressure charging pump refilling the PCS to a 

point approaching water-solid conditions) if PCS makeup were initiated without establishing a 

vent path. 

However, the NRG staff observed that the licensee has the capability to vent the PCS if needed, 

via primary coolant gas vent system vent paths from both the reactor vessel upper head and 

pressurizer. These vent paths contain 7/32-in. orifices that limit the rate at which PCS coolant 

can be let down, and the valves in the flowpath are operated by de power. Furthermore, as 

discussed during the audit, the licensee's FLEX procedures direct that a PCS vent path be 

opened when FLEX PCS makeup is being performed. According to procedure, the vent path 

would subsequently be isolated when FLEX PCS makeup is secured. Therefore, the NRG staff 

concluded that the licensee has an appropriate strategy for venting the PCS under ELAP 

conditions. Considering the limited volume of BAST injection required to ensure adequate 

shutdown margin and the size of the vent line orifice, in light of the results for the more limiting 

case documented in the licensee's shutdown margin calculation, the NRG staff's audit review 

concluded that the necessary actions could be completed within the required timeframe. 
- 24 -

Guidance document NEI 12-06, Section 11.8.2, states that plant configuration control 

procedures will be modified to ensure that changes to the plant design will not adversely impact 

the approved FLEX strategies. Inasmuch as changes to the core design constitute changes to 

the plant design, the staff expects that any changes to the core design, such as those evaluated 

in a typical core reload analysis, will be evaluated to determine that they do not adversely 

impact the approved FLEX strategies, especially the analyses which demonstrate that no recriticality will occur during a FLEX PCS cooldown. 

Therefore, based on the evaluation above, the NRC staff concludes that the sequence of events 

in the proposed mitigating strategy should result in acceptable shutdown margin for the 

analyzed ELAP event. 

3.2.3.5 FLEX Pumps and Water Supplies 

The licensee's Phase 2 FLEX strategy relies on a centrifugal, trailer-mounted, portable diesel 

driven FLEX pump that can provide core cooling, SFP cooling and boric acid batching water 

supply. This FLEX pump takes suction from the UHS and discussion of its robustness is 

documented in SE Section 3.10. The licensee identified the performance criteria (e.g., flow rate, 

discharge pressure, total dynamic head) for its FLEX pump in FIP Section 3.2.10, which 

describes a FLEX pump supplied from a NSRC for Phase 3 would allow the NSRC pump to 

fulfill the mission of the onsite FLEX pump if the onsite FLEX pump were to fail. 

The FLEX pump takes suction from the intake canal through removable grating and discharges 

to a portable FLEX manifold. From this manifold, hoses can be routed to the MFW system 

(primary connection) or the AFW system (alternate connection), the SFP and the boric acid 

tank. A single diesel driven FLEX pump provides full capability to perform all the FLEX 

functions with one portable pump stored in each of the two FLEX Storage Buildings; thus, the 

licensee satisfies the N+ 1 requirement outlined in NEI 12-06. 

During the audit, the licensee provided Calculation EC-46465-06, "Hydraulic Analysis of FLEX 

Coping Strategies," Rev. 0, which determined the necessary pump performance criteria for the 

portable diesel driven FLEX pumps to support the licensee's core cooling, SFP cooling and 

boron batching strategies. The staff noted that this calculation assessed numerous possible 

lineups based on such variables as suction sources, connection points and hose paths to 

determine adequate performance criteria for the portable diesel driven FLEX pump while 

simultaneously injecting into the SGs, providing SFP makeup and providing water for boron 

batching. The calculation determined the total flow requirement was 410.5 gpm at 235 psi 

discharge pressure broken down to the following: 136.5 gpm into the SGs, 250 gpm of SFP 

overspray, 15 gpm for boron batching and 8.5 gpm margin. The FLEX pump can provide 41 O 

gpm at 275 psi. 

During the audit, the NRC staff performed a walkdown of the licensee's core cooling FLEX 

strategies and noted that the point of deployment for the portable FLEX pumps, hose routing 

and deployment connection points (primary and alternate) were consistent with the licensee's 

hydraulic analysis. The staff noted that the staging and operation of the portable diesel driven 

FLEX pump are directed in procedure, FIG-2, "FLEX Pump Stating and Operation," Rev. 0, and 

that procedure FIG-4, "Low Pressure Feedwater Alignment," Rev. 0, directs operators to feed 

the SGs using the portable diesel driven FLEX pump. 
- 25 -

The licensee's Phase 2 RCS inventory control strategies rely on the use of one of two installed 

positive displacement charging Pumps taking suction from the BASTs to support injection into 

the RCS through the charging pumps discharge header. The licensee stated in its FIP that 32 

gpm of RCS makeup is required and the each charging pump can provide 44 gpm. The 

licensee does plan to use a hose connected to the FLEX manifold to provide water from the 

portable diesel driven FLEX pump for boric acid mixing. The installed charging pumps were 

discussed further in Section 3.2.3.1.1. Repowering installed charging pumps to provide RCS 

makeup is considered an alternative to NEI 12-06 and acceptability of this alternative is 

documented in SE Section 3.14.1. 

The FIP states they will continue with Phase 2 strategies during Phase 3 using the mobile 

boration and water purification units to purify and provide clean water from Lake Michigan. 

However, the licensee does intend on establishing SOC strategies that rely on the following 

pumps provided by the NSRC: a diesel-driven, low pressure, high flow, pump (5000 gpm and 

150 psi) to provide flow to the SWS system, and a hydraulically driven, booster pump (26 ft. 

water lift and 5000 gpm) to provide flow to the high flow/low pressure pump. These pumps from 

the NSRC would be used in conjunction with the repowered CCW pump and SDC/LPSI pump to 

establish SOC. During the audit, the staff noted the licensee's procedure, FIG-11,"Service 

Water - NSRC (Phase 3)," provides guidance for deploying the NSRC pump, the booster pump, 

and the associated SWS strainer adapter, hoses, manifolds, and fittings. 

Based on the staff's review of the FLEX pumping capabilities at PNP, as described in the above 

hydraulic analyses and the FIP, the NRC staff concludes that the portable FLEX pumps should 

perform as intended to support core cooling and RCS inventory control during an ELAP event, 

consistent with NEI 12-06, Section 11.2. 

3.2.3.6 Electrical Analyses 

The licensee's FIP defined strategies capable of mitigating a simultaneous loss of all ac power 

and loss of normal access to the ultimate heat sink (LUHS) resulting from a BDBEE by providing 

the capability to maintain or restore core cooling at Palisades. The Palisades electrical FLEX 

strategies are practically identical for maintaining or restoring core cooling, containment, and 

spent fuel pool cooling, except as noted in Sections 3.3.4.4 and 3.4.4.4 of this SE. Furthermore, 

the electrical coping strategies are the same for all modes of operation. 

The NRC staff reviewed the licensee's FIP to determine whether the FLEX strategies, if 

implemented appropriately, should maintain or restore core cooling, containment, and spent fuel 

pool cooling following a BDBEE. As part of its review, the NRC staff reviewed conceptual 

electrical single-line diagrams, summaries of calculations for sizing the FLEX diesel and turbine 

generators and station batteries, and summaries of calculations that addressed the effects of 

temperature on the electrical equipment credited in the FIP as a result of losing heating, 

ventilation, and air conditioning (HVAC) during an ELAP as a result of a BDBEE. 

According to the licensee's FIP, the Palisades Operators would declare an ELAP following a 

loss of offsite power, loss of all emergency diesel generators, and the loss of any alternate ac 

power with a simultaneous loss of normal access to the LUHS. In the FIP, the licensee 

assumes that this determination can be made less than1 hour after the onset of an ELAP/LUHS 

event. 
- 26 -

During Phase 1, Palisades would rely on the station's safety-related batteries to provide support 

to instrumentation for monitoring parameters and support to controls for SSCs used to maintain 

the key safety functions (Reactor core cooling, PCS inventory control, and Containment 

integrity). The Palisades station batteries (ED-01 and ED-02) are safety-related Class 1 E SSCs 

that are fully protected within the Auxiliary Building (Seismic Category 1 structure) and fully 

qualified for all applicable external events. The safety-related batteries were manufactured by 

C&D Technologies and are model LCR-25 that are rated 1770 ampere-hour at an 8-hour 

discharge rate to 1. 78 volts per cell. Following the reactor transient as a result of losing offsite 

power and the unavailability of the emergency diesel generators, the licensee will initially shed 

loads from the battery utilizing the existing SBO procedure. Once an ELAP has been declared 

(approximately one hour after the event when it becomes evident that ac power will not be 

restored from onsite or off-site sources), the Palisades operators would initiate procedures to 

shed additional unnecessary loads on the station's safety-related batteries (FSG-4, "ELAP DC 

Load Shed and Management," Rev. 0). The licensee expects load shedding to be completed 

within 2 hours of event initiation. Based on its review of calculation EA-ELEC-LDTAB-021, 

"Station Batteries ED-01 & ED-02 FLEX Coping Capability," Rev. 0, the NRG staff confirmed 

that the Palisades station safety-related batteries will be available for a minimum of 8 hours 

during Phase 1 provided the load shedding is completed within 2 hours of event initiation. The 

calculated available battery capacity is sufficient since the licensee expects to have a FLEX 480 

Vac Diesel Generator deployed, staged, and connected as part of its phase 2 transition to 

repower a station 480 Vac bus to power battery chargers within 8 hours of the onset of an 

ELAP/LUHS event. 

Based on the evaluation above, the NRG staff concludes that the PNP load shed strategy 

should ensure that the batteries have sufficient capacity to supply power to required loads until 

the portable generators are deployed. 

During Phase 2, the licensee will deploy one portable 480 Vac 463 kW FLEX diesel generator to 

restore power to the Class 1 E Battery Charger, power up a Charging Pump, flow path motoroperator valves (MOVs), emergency lighting, the electric driven FLEX air compressor and the 

electric driven FLEX portable Battery Room ventilation. Two 480 Vac 463 kW FLEX diesel 

generators are available to satisfy N+ 1 requirement. The licensee expects that plant operators 

will begin deployment of a 463 kW 480 Vac FLEX portable diesel generator within 2 hours of 

event initiation to restore power to LCC-19 or LCC-20 and other strategy loads. The licensee 

plans to complete this action within 8 hours of event initiation to prevent station battery (ED-01 

and ED-02) depletion. Once LCC-19 (primary connection point) or LCC-20 (alternate 

connection point) is being supplied power from the portable diesel driven 463 kW 480 Vac FLEX 

generator, MCC-1 or MCC-2 can be energized, which allows 2 of 4 Battery Chargers to be 

placed in service to recharge the station batteries using the normal Class 1 E breakers. 

The FLEX diesel generator will also provide power to the electric driven portable ventilation fan 

to remove hydrogen from the Battery Rooms, and provide the source of power to the portable 

electric driven air compressor which will supply air to the ADVs in support of SG pressure 

control and PCS temperature control. Once the Battery Chargers are placed in service, a 

portable electric driven FLEX ventilation fan will be placed in service for the associated Battery 

Room to maintain hydrogen below combustibility limits. To ensure hydrogen limits remain 

below combustibility limits, the portable electric driven FLEX ventilation/exhaust fan must be 

placed in service in the Battery Room by 1.5 hours after battery charging begins, or 9.5 hours 
- 27 -

after event initiation. The licensee will route a flexible duct from the Cable Spreading Room 

outside of the Turbine Building such that the hydrogen can be exhausted to the atmosphere, 

preventing hydrogen accumulation in other areas of the plant. 

At the point when ELAP mitigation activities require connection of the FLEX diesel generators, 

in addition to existing electrical interlocks, the licensee will employ procedural controls, such as 

inhibiting EOG start circuits and breaker rack-outs (e.g., EOG breakers, offsite feeder breakers, 

etc.), to prevent simultaneous connection of both the FLEX diesel generators and Class 1 E 

EDGs to the same ac distribution system or component. Additionally, repowering the Class 1 E 

electrical buses from either FLEX diesel generators or subsequently the Class 1 E EDGs (should 

they become available) will be accomplished manually and controlled by procedure; no 

automatic sequencing or automatic repowering of the buses will be utilized. 

The NRG staff reviewed the licensee's Phase 2 and 3 FLEX generator calculations and 

procedures (EA-ELEC-LDTAB-022, "PLP FLEX Diesel Sizing Calculation," Rev. 0, Attachment 

9.3.14 of Engineering Change 46465 (Phase 3 4160 Vac Turbine Generator sizing), FSG-5, 

"Initial Assessment and FLEX Equipment Staging," Rev. 0, FIG-1, "FLEX Generator Staging and 

Operation," Rev. 0, and FIG-12, "4160 Vac Generator - NSRC (Phase 3)," Rev. 2). Calculation 

EA-ELEC-LDTAB-022, "PLP FLEX Diesel Sizing Calculation," Rev. 0, showed that the required 

Phase 2 loads equal 325.11 kW. Therefore, one 463 kW FLEX Diesel Generator is adequate to 

support the electrical loads required for the licensee's Phase 2 strategies. 

For Phase 3, the licensee will receive two 1-MW, 4160 Vac, 3-phase turbine generators 

(including a distribution panel, to connect them to a plant supplied FLEX 4160/2400 Vac 

transformer) and one 1-MW, 480 Vac, 3-phase turbine generator from an NSRC. These turbine 

generators could be used to supply power battery chargers, ventilation, MOVs, and other 

miscellaneous loads. The NRG staff reviewed calculations, conceptual single line electrical 

diagrams, the separation and isolation of the FLEX turbine generators from the Class 1 E EDGs, 

and procedures that direct operators how to align, connect, and protect associated systems and 

components. Each turbine generator is capable of supplying approximately 1 MW, but two 

turbine generators will be operated in parallel to provide approximately 2 MW (2.5 MVA at .8 pf). 

Per calculation EA-ELEC-LDTAB-022, "PLP FLEX Diesel Sizing Calculation," Rev. 0, the total 

loads for Phase 3 equal 1646.24 kW. 

The two 1-MW 4160 Vac turbine generators will provide the ability to supply power to one 

complete train of Class 1 E equipment that can be used to bring the plant to cold shutdown and 

restore other auxiliaries such as CACs. One bus of Class 1 E 2400 V switchgear will be 

energized with the NSRC turbine generator. Specifically, the 4160 Vac turbine generator will be 

used to supply (via a distribution panel and transformer) 2400 Vac power to Bus 1 C or Bus 1 D. 

The primary connection point is Bus 1 D, due to its close proximity to where the NSRC turbine 

generator will be staged. Bus 1 C is an alternate connection point for powering redundant loads 

in case power through Bus 1 D fails or is unable to be used. Because the turbine generator will 

supply 4160 Vac power, a 4160/2400 V transformer is required to step down the voltage. The 

4160/2400 V transformer will be connected to Bus 1 C or 1 D using temporary cables connected 

directly to the bus bars (all three phases) of an existing breaker 152-213 on Bus 1 D (or 152-107 

on Bus 1 C). No breaker modifications are needed, but the connections will replace those from 

the onsite EOG for that bus. The transformer neutral will be grounded on the 4160 V side, and 

ungrounded on the 2400 V side, to match the system. The 4160 Vac NSRC supplied turbine 

generators have been sized to ensure the capability (as a minimum) to energize a CCW Pump, 
- 28 -

LPSI Pump, SOC flowpath MOVs, a SFP Pump, and 3 CACs. Additionally, the licensee has 

sized the plant supplied Phase 3 FLEX step down transformer to accommodate the capacity of 

2 NSRC 1-MW turbine generators. Power cables will be supplied with the NSRC 4160 Vac 

turbine generators for connection to the Palisades Plant electrical buses. 

The 480 Vac NSRC turbine generators come with the same style and size connectors as the 

on-site Phase 2 FLEX generators. Therefore, the licensee's Phase 3 strategy could utilize the 

Phase 2 electrical connections if needed. The capacity of the NSRC-supplied 480 Vac turbine 

generator is of greater capacity than the capacity of the licensee's Phase 2 FLEX Diesel 

Generators. Therefore, the NRG staff finds that the Phase 3 480 Vac turbine generators will 

provide adequate capacity to supply electrical loads (same as Phase 2) to maintain or restore 

core cooling, SFP cooling, and containment indefinitely following an ELAP. 

Based on its review, the NRG staff finds that the plant batteries used in the strategy should have 

sufficient capacity to support the licensee's strategy, and that the FLEX DGs and turbine 

generators that the licensee plans to use should have sufficient capacity and capability to supply 

the necessary loads during an ELAP event. 

3.2.4 Conclusions 

Based on this evaluation, the NRG staff concludes that the licensee has developed guidance 

that should maintain or restore core cooling during an ELAP event consistent with NEI 12-06 

guidance, as endorsed by JLD-ISG-2012-01, with the alternatives described in Section 3.14 of 

this SE, and adequately addresses the requirements of the order. 

3.3 Spent Fuel Pool Cooling Strategies 

In NEI 12-06, Table 3-2 and Appendix D, summarize an acceptable approach consisting of 

three separate capabilities for the SFP cooling strategies. This approach uses a portable 

injection source to provide the capability for 1) makeup via hoses on the refueling floor capable 

of exceeding the boil-off rate for the design-basis heat load; 2) makeup via connection to spent 

fuel pool cooling piping or other alternate location capable of exceeding the boil-off rate for the 

design basis heat load; and 3) spray via 2 portable monitor nozzles from the refueling floor 

using a portable pump capable of providing 125 gpm per nozzle or 250 gpm total. During the 

event, the licensee selects the method to use based on plant conditions. This approach also 

requires a strategy to mitigate the effects of steam from the SFP, such as venting. 

As described in NEI 12-06, Section 3.2.1. 7 and JLD-ISG-2012-01, Section 2.1, strategies that 

must be completed within a certain period of time should be identified and a basis that the time 

can be reasonably met should be provided. In NEI 12-06, Section 3 provides the performance 

attributes, general criteria, and baseline assumptions to be used in developing the technical 

basis for the time constraints. Since the event is beyond-design-basis, the analysis used to 

provide the technical basis for time constraints for the mitigation strategies may use nominal 

initial values (without uncertainties) for plant parameters, and best-estimate physics data. All 

equipment used for consequence mitigation may be assumed to operate at nominal setpoints 

and capacities. In NEI 12-06, Section 3.2.1.2 describes the initial plant conditions for the atpower mode of operation; Section 3.2.1.3 describes the initial conditions; and Section 3.2.1.6 

describes SFP initial conditions. 
- 29 -

In NEI 12-06, Section 3.2.1.1, provides the acceptance criterion for the analyses serving as the 

technical basis for establishing the time constraints for the baseline coping capabilities to 

maintain SFP cooling. This criterion is keeping the fuel in the SFP covered with water. 

The ELAP causes a loss of cooling in the SFP. As a result, the pool water will heat up and 

eventually boil off. The licensee's response is to provide makeup water. The timing of operator 

actions and the required makeup rates depend on the decay heat level of the fuel assemblies in 

the SFP. The sections below address the response during operating, pre-fuel transfer or postfuel transfer operations. The effects of an ELAP with full core offload to the SFP is addressed in 

Section 3.11 . 

3.3.1 Phase 1 

The FIP states no action is required for Phase 1 makeup because the time to uncover the fuel is 

sufficient to deploy Phase 2 equipment. However, the FIP states that boiling may start as soon 

as 3.28 hours during an outage with the entire core located in the SFP after the ELAP (5.63 

hours with a normal, non-outage SFP heatload), and may affect personnel habitability on the 

refueling floor. For this reason, the licensee's strategy is to complete the hose deployment and 

establish a vent path on the refueling floor prior to 3 hours after an ELAP declaration. The FIP 

also states that the time for boil-off to lower the water level to the top of the fuel racks is 11.07 

hours after the loss of normal SFP cooling with the conservative assumption of a full-core 

offload scenario. 

3.3.2 Phase 2 

The licensee's Phase 2 SFP cooling strategy consists of deploying the portable diesel driven 

FLEX pump to provide makeup from Lake Michigan. As described in the FIP, hoses can be 

connected to the FLEX pump discharge manifold and routed to the refuel floor to provide direct 

makeup to the pool and spray flow. Additionally, a hose can be connected to the installed SFP 

cooling system, which does not require access to the refueling floor. As described in Section 

3.2.3.5, the FLEX manifold is pressurized using the portable FLEX pump staged at the intake 

structure. The Diesel Driven FLEX Pump is capable of providing up to 250 gpm of spray flow. 

3.3.3 Phase 3 

The FIP states that the Phase 3 strategy can continue to utilize the Phase 2 strategies. 

However, the licensee will establish CCW and SWS, as described in SE Section 3.2.1 .2.3. The 

licensee would then be able to restore SFP Cooling system operation. 

3.3.4 Staff Evaluations 

3.3.4.1 Availability of Structures. Systems. and Components 

3.3.4.1.1 Plant SSCs 

Condition 6 of NEI 12-06, Section 3.2.1.3, states that permanent plant equipment contained in 

structures with designs that are robust with respect to seismic events, floods, and high winds, 

and associated missiles, are available. In addition, Section 3.2.1.6 states that the initial SFP 

conditions are: 1) all boundaries of the SFP are intact, including the liner, gates, transfer canals, 
- 30 -

etc., 2) although sloshing may occur during a seismic event, the initial loss of SFP inventory 

does not preclude access to the refueling deck around the pool and 3) SFP cooling system is 

intact, including attached piping. 

The FIP indicates that boiling begins at approximately 5.63 hours during a normal, non-outage 

situation. This analysis assumes a heat load of 4.47 million British Thermal Unit (Btu)/hr and a 

starting SFP temperature of 140°F. The staff noted that the licensee's sequence of events 

timeline in the FIP indicates that operators will deploy hoses and spray nozzles as a 

contingency for SFP makeup within 3 hours from event initiation. The staff noted that 

Procedure FIG-3, "Alternate SFP Makeup and Cooling,'' Rev. 0, contains a precaution that SFP 

hoses should be deployed prior to 3 hours to ensure the SFP area remains habitable for 

personnel entry. 

As described in the FIP, the Phase 1 SFP cooling strategy does not require any prestaged 

equipment. However, the licensee does establish a ventilation path to cope with temperature, 

humidity and condensation from evaporation and/or boiling of the SFP. The staff noted that 

Procedure FIG-8, "Alternate Ventilation,'' Rev. 0, contains instructions for operators to establish 

a SFP vent path following the ELAP by opening the double leaf doors on the roof of the Fuel 

Building. 

The licensee's Phase 2 and Phase 3 SFP cooling strategy involves use of the FLEX pump (or 

NSRC supplied pump for Phase 3), with suction from the UHS, to pressurize the portable FLEX 

manifold. The portable FLEX manifold can then be connected via hose to provide direct 

makeup and spray flow or can be connected to the installed SFP cooling system piping. The 

staff's evaluation of the robustness and availability of the UHS for an ELAP event is discussed 

in Section 3.10.3. 

3.3.4.1.2 Plant Instrumentation 

In its FIP, the licensee stated that the instrumentation for SFP level will meet the requirements 

of Order EA-12-051. Furthermore, the licensee stated that these instruments will have initial 

local battery power with the capability to be powered from the FLEX DGs. The NRC staff's 

review of the SFP level instrumentation, including the primary and back-up channels, the display 

to monitor the SFP water level and environmental qualifications to operate reliably for an 

extended period are discussed in Section 4 of this safety evaluation. 

3.3.4.2 Thermal-Hydraulic Analyses 

In its FIP, the licensee analyzed two different SFP cooling heat load cases when determining 

SFP makeup rate. These two cases are a maximum normal heat load during an outage 

following a normal batch discharge (Case 1 ), and a maximum abnormal heat load during an 

outage following a full core discharge (Case 2). The maximum heat load, boil-off time to top of 

the fuel and makeup rate can be found in the table below. 

Heat Load Time to boil to 15 ft Makeup rate 

from top of fuel 

Case 1 4.47 million Btu/hr 21.3 hrs 33gpm 

Case 2 9.02 million Btu/hr 11.07 hrs 65gpm 
- 31 -

As stated in SE Section 3.2.3.5, the licensee's FLEX Pump is capable of supplying up to 250 

gpm of SFP spray flow, which is more than either worst case SFP makeup requirements. 

The staff noted that NEI 12-06, Section 3.2.1.6, states that one of the initial SFP conditions is 

that the SFP heat load assumes the maximum design-basis heat load for the site. Consistent 

with NEI 12-06, Section 3.2.1.6, the staff finds that the licensee has considered the maximum 

design-basis SFP heat load. Considering that the ELAP event is postulated to start from 100 

percent power, it is not possible to have a full core offload in the SFP, and FIP states the 

required makeup rate would be 33 gpm. The licensee has the capability to provide this makeup 

rate. 

3.3.4.3 FLEX Pumps and Water Supplies 

As described in the FIP, the SFP cooling strategy relies on a FLEX pump to provide SFP 

makeup during Phase 2. In the FIP, Section 3.2.10 describes the hydraulic performance criteria 

(e.g., flow rate, discharge pressure) for the FLEX Pump. The staff noted that the performance 

criteria of a FLEX pump supplied from an NSRC for Phase 3 would allow the NSRC pump to 

fulfill the mission of the onsite FLEX pump if the onsite FLEX pump were to fail. As stated 

above, the SFP makeup rate of 33 gpm and SFP spray rate of 250 gpm both meet or exceed 

the maximum SFP makeup requirements as outlined in the previous section of this SE. Section 

3.2.3.5 in the SE discusses the performance criteria (e.g., flow rate, discharge pressure, total 

dynamic head) of the portable diesel driven FLEX pump. Additionally during the onsite audit, 

the NRG staff performed a walkdown of the licensee's SFP cooling FLEX strategies and noted 

that the staging location for the portable FLEX pumps, hose routing and connection points were 

consistent with the licensee's hydraulic analysis. Further discussion regarding the robustness 

and availability of these connections and water sources are documented in SE Sections 3.7.3.1 

and 3.10.3, respectively. 

3.3.4.4 Electrical Analyses 

The licensee's OIP and FIP define strategies capable of mitigating a simultaneous ELAP and 

loss of normal access to the ultimate heat sink resulting from a BDBEE by providing the 

capability to maintain or restore core cooling, containment, and SFP cooling at the PNP site. 

The electrical coping strategies are the same for all modes of operation. 

The NRG staff performed a comprehensive analysis of the licensee's electrical strategies, which 

includes the SFP cooling strategy. The NRG staff's review is discussed in detail in Section 

3.2.3.6. 

3.3.5 Conclusions 

Based on this evaluation, the NRG staff concludes that the licensee has developed guidance 

that, if implemented appropriately, should maintain or restore SFP cooling following an ELAP 

consistent with NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01, and should adequately 

address the requirements of the order. 
- 32 -

3.4 Containment Function Strategies 

The industry guidance document, NEI 12-06, Table 3-2 provides some examples of acceptable 

approaches for demonstrating the baseline capability of the containment strategies to effectively 

maintain containment functions during all phases of an ELAP event. One such approach is for a 

licensee to perform an analysis demonstrating that containment pressure control is not 

challenged. 

In accordance with NEI 12-06, Palisades performed a containment evaluation EA-EC46465-02, 

"MAAP Containment Analysis for BDBEE," Rev. 0, which was based on the boundary conditions 

described in Section 2 of NEI 12-06. The calculation concludes that the containment 

parameters of pressure and temperature remain well below the respective design limits of 55 

psig and 283°F (UFSAR Section 5.8) for at least 1 O days. In addition, key parameters 

instrumentation subject to the containment environment will remain available. From its review of 

the evaluation, the NRG staff noted that the required actions to maintain containment integrity 

and required instrumentation functions have been developed, and are summarized below. 

The NRG staff reviewed calculation EA-EC46465-02, "MAAP Containment Analysis for 

BDBEE," Rev. 0. The calculation used the Modular Accident Analyses Program (MAAP) 4.0.6 

computer code. The calculation forecasts the Containment pressure and temperature 

(pressures and temperatures are highest in the reactor cavity compartment) initially rise on loss 

of normal containment cooling to approximately 4.5 psig and 205°F, respectively. After the first 

8 hours, the pressure and temperature gradually increases at a steady state. The maximum 

Containment pressure reached was calculated to be 8.6 psig (at 120 hours), remaining well 

below the design-basis limit of 55 psig. The calculation also indicates the maximum 

Containment temperature reached was 218.2°F (at 120 hours), which remains below the design 

limit of 283°F. 

Eventual containment cooling and depressurization to normal values may utilize off-site 

equipment and resources during Phase 3. For the at-power scenario with SGs removing core 

heat, no specific coping strategy is required for maintaining containment integrity during Phase 

1, 2 or 3. Results of the Mode 5 scenario at mid-loop inventory, with containment venting is 

discussed below in Section 3.11. 

3.4. 1 Phase 1 

Containment pressure and temperature are expected to increase during an ELAP due to loss of 

Containment cooling and mass and energy transfer from the primary coolant system (PCS) to 

Containment. The Palisades Containment design pressure is 55 psig, and the design 

temperature is 283°F. The Containment Liner has been analyzed for temperatures up to 410°F 

for a design basis accident. 

A Containment evaluation has been performed consistent with the boundary conditions 

described in Section 2 of NEI 12-06. Based on the performance of installed primary coolant 

pump (PCP) seals, the pressure and temperature of Containment are not expected to rise 

significantly. Calculation CN-SEE-11-13-5, "Palisades Reactor Coolant System (RCS) Inventory 

and Shutdown Margin Analysis to Support the Diverse and Flexible Coping Strategies (FLEX)," 

Rev. 1, demonstrated that as long as cooling water was restored to the SGs prior to fuel 

damage there would not be any structural concerns with Containment for this event. 
- 33 -

The Phase 1 coping strategy for containment involves verifying containment isolation per 

procedure EOP-3.0, "Station Blackout Recovery," and monitoring containment temperature and 

pressure using installed instrumentation. 

The results of the MAAP analysis confirm that the Containment function is not challenged early 

in the event for the at-power scenario. Therefore, no Phase 1 actions are required to maintain 

Containment integrity. 

3.4.2 Phase 2 

Consistent with WCAP-17601-P, "Reactor Coolant System Response to the Extended Loss of 

AC Power Event for Westinghouse, Combustion Engineering and Babcock and Wilcox NSSS 

Designs," Rev. 0, the Mode 1 analysis assumes an initial letdown rate of 75 gpm for the first ten 

minutes of the transient, and conservatively assumes an initial PCP seal leakage rate of 15 gpm 

per seal (PCP seal leakage decreases as the PCS is depressurized). The Containment 

analysis assumed a PCP seal leakage rate of 15 gpm because this maximizes the energy 

transfer into Containment and is therefore conservative. The analysis for Mode 1 follows the 

75°F per hour cooldown case documented in the Westinghouse PCS analysis CN-SEE-11-13-5. 

The Containment analysis, EA-EC46465-02, determines Containment temperature and 

pressure for the analysis duration of 120 hours. Results of this analysis show that Containment 

temperature and pressure remain below their respective design limits for the analysis duration 

of 120 hours. The rate of temperature and pressure increase in Containment at 120 hours is 

such that the design limits will not be challenged within the first 1 O days after an event. Prior to 

120 hours after the event, onsite and off-site equipment will be available to mitigate 

Containment conditions. 

The Palisades containment analysis shows that there are no Phase 2 actions required. 

Containment pressure and temperature will continue to be monitored using installed 

instrumentation. 

3.4.3 Phase 3 

The containment analysis, EA-EC46465-02, showed that the CACs are not required before 120 

hours for the at-power scenario. During the plant recovery phase (i.e. Phase 3) the CACs may 

be utilized, although they are not credited for the FLEX coping strategy. 

The NSRC will supply a low pressure/high flow pump and two 4160 Vac generators for longterm plant recovery. The low pressure/high flow pump allows for the restoration of service water 

flow to equipment such as the CCW system and LPSI pump. The 4160 Vac generators allow 

for the restoration of one complete Class 1 E distribution bus (1 D or 1 C) which provide power to 

a LPSI pump, a CCW pump, SOC flowpath valves and 3 CACs. 

The licensee's analysis concluded that containment temperatures will remain below acceptable 

limits for critical instrumentation inside Containment for at least 6 days. The NSRC generator 

will be utilized to start and stop CACs as required to maintain temperatures within limits to 

preserve critical instrumentation. As required, the NSRC low pressure/high flow pump and 
- 34 -

NSRC 4160 Vac turbine generators can be used to transition the plant to a long term recovery 

mode utilizing SOC, thereby maintaining Containment temperatures and pressures below 

design limits. 

3.4.4 Staff Evaluations 

3.4.4.1 Availability of Structures, Systems, and Components 

The NEI 12-06 baseline assumptions have been established on the presumption that other than 

the loss of the ac power sources and normal access to the UHS, installed equipment that is 

designed to be robust with respect to design-basis external events is assumed to be fully 

available. Installed equipment that is not robust is assumed to be unavailable. Below are the 

baseline assumptions for the availability of SSCs for maintaining Containment functions during 

an ELAP. 

3.4.4.1.1 Plant SSCs 

Containment Building 

The containment building is a post-tensioned concrete cylinder and dome connected to a 

conventionally reinforced slab foundation. Its entire inner surface is lined with a welded steel 

plate to ensure a high degree of leak tightness. The containment building completely encloses 

the reactor and the Primary Coolant System. The Containment Building is a Design Class 1, 

Seismic Category 1 structure that is qualified for all applicable external events. 

Containment Air Coolers 

The CACs, ducting and associated Service Water Valves are safety-related, Design Class 1, 

Seismic Category 1 structures. All equipment is located within the Design Class 1, Seismic 

Category 1 Containment Building. All equipment is fully protected and qualified for all applicable 

external events. 

Service Water System 

The SWS piping will provide the flowpath for cooling water from the NSRC supplied low 

pressure/high flow pump for cooling of various plant loads. Pressure boundary must be ensured 

to credit adequate flow. The service water connection piping is located within the Design Class 

1 portion of the Intake Structure. The SWS piping segments utilized to support safety-related 

loads for plant recovery actions (CCW heat exchangers, CACs and LPSI pump via CCW) are 

supplied by the Design Class 1 SWS "A" or "B" critical headers. As such, these portions of the 

system are fully protected and qualified for all applicable external events. 

CCW System 

Portions of the CCW system utilized in the recovery phase of the Phase 3 FLEX strategy are 

located outside Containment within the Auxiliary Building. The Auxiliary Building is a Design 

Class 1 structure that fully protects the equipment from all applicable external hazards. The 

portions of the system piping outside Containment are safety-related, Seismic Category 1 

structures. The CCW pumps are supplied power from Class 1 E safety-related switchgear and 
- 35 -

associated cabling/controls. The CCW pumps themselves are seismically qualified, safetyrelated SSCs. Based on the above, all CCW SSCs required for the Phase 3 FLEX strategy are 

fully protected and qualified for all applicable external events. 

LPSI System 

The LPSI pump, associated piping and SOC heat exchanger provide the SOC function for the 

station. These components are used in the recovery phase of Phase 3. The associated piping 

segments and SOC heat exchanger of LPSI are Design Class 1 per the UFSAR Table 5.2-3 and 

located inside the Auxiliary Building such that the SSCs are protected from wind and flooding 

external events. Per UFSAR Table 5.2.3, interconnecting valves required to perform 

recirculation function (including HPSI and LPSI systems) are Class 1 safety-related. The MOVs 

are powered by Class 1 E MCC-1 and, therefore, the MOV motors are completely qualified since 

they are bounded by the Class 1 SSCs they are connected to. Likewise, the LPSI pump (SOC 

pump) is Design Class 1 with Class 1 E powered motor, located within Auxiliary Building and is 

fully qualified for all applicable external events. The subject SSCs supporting the FLEX strategy 

are fully protected and qualified for all applicable external events. 

3.4.4.1.2 Plant Instrumentation 

In NEI 12-06, Table 3-2, specifies that containment pressure is a key containment parameter 

which should be monitored by repowering the appropriate instruments. In the Palisades FIP, 

the licensee stated that the key parameters for the containment integrity function are 

containment pressure and containment temperature, which can be obtained from essential 

instrumentation (LPIR-0382/0383; and Tl-1812, Tl1813, Tl1814, and Tl1815, respectively). 

The above instrumentation are seismically robust, fed by Class 1 E power and are located within 

the Seismic Category 1 Auxiliary Building/Containment Building/MGR and as such, are qualified 

and protected for all applicable external events. The instruments indication are available (at 

least one channel depending on which train of load shed is performed) in the MCR for all three 

FLEX strategy phases (1, 2, and 3). 

In the unlikely event that bus infrastructure is damaged, alternate FLEX strategy guidelines for 

obtaining the critical parameters locally is provided in FSG-7, "Loss of Vital Instrument or 

Control Power," in accordance with the guidelines of NEI 12-06 Section 5.3.3. 

3.4.4.2 Thermal-Hydraulic Analyses 

The NRC staff reviewed calculation EA-EC46465-02, "MAAP Containment Analysis for 

BDBEE," Rev. 0. This calculation used the MAAP computer code, Version 4.0.6. The analysis 

evaluates the containment response with the plant initially at full power in Mode 1 at the time of 

a BDBEE, and bounds plant configurations in Modes 1 through 4 with SGs available. The 

volume of water used to feed the SGs is approximately 576,000 gallons. The volume of borated 

water added to the primary system with a charging pump is approximately 43,600 gallons. An 

initial containment pressure was assumed to be O psig and an initial containment temperature 

was assumed to be 140°F. The calculation modeled the containment pressure and temperature 

over a 120-hour period during an extended loss of ac power event. 
- 36 -

The TOAFW pump is used to feed the SGs with suction from the condensate storage tank for 

the first 24 hours. The steam extraction rate and injection flow rate is split evenly between both 

SGs. At 8 hours, SG injection is switched to the portable FLEX injection pumps. The portable 

FLEX pump will have a capacity of 230 gpm, and feed will be split evenly between the SGs. 

The assumed reactor coolant leakage is 15 gpm per RCP and 1 gpm from unidentified sources 

for a total leakage of 61 gpm. 

The maximum Containment pressure reached was calculated to be 8.6 psig (at 120 hours) and 

the maximum Containment temperature reached was 218.2°F (at 120 hours). The Containment 

building remains below the internal design limits of 55 psig and 283°F. 

3.4.4.3 FLEX Pumps and Water Supplies 

The NSRC is supplying a low pressure/high flow FLEX pump which will be connected to the 

SWS to support operation of SOC as described in Section 3.2.3.5 of this SE. Water supply will 

be provided by the Intake Canal/Lake Michigan. 

3.4.4.4 Electrical Analyses 

The licensee performed a containment evaluation analysis based on the boundary conditions 

described in Section 2 of NEI 12-06. Based on the results of this analysis, the licensee 

developed required actions to ensure maintenance of containment integrity and required 

instrumentation function. With an ELAP initiated, while Palisades is in Modes 1-4, containment 

cooling for that unit is also lost for an extended period of time. Therefore, containment 

temperature and pressure will slowly increase. The licensee's analysis concluded that 

containment temperature and pressure will remain below containment design limits and that 

essential instruments subject to the containment environment will remain functional for a 

minimum of 6 days. Therefore, actions to reduce containment temperature and pressure and to 

ensure continued functionality of the key parameters will not be required immediately. Eventual 

containment cooling and depressurization to normal values may utilize off-site equipment and 

resources during Phase 3. 

The Phase 1 coping strategy for containment involves verifying containment isolation per 1 (2)-

EOP-3.0, "Station Blackout Recovery," and monitoring containment temperature and pressure 

using installed instrumentation. The Phase 2 coping strategy is to continue monitoring 

containment temperature and pressure using installed instrumentation. The Phase 3 coping 

strategy discussed in Section 3.2.3 of the FIP is to obtain additional electrical capability and 

redundancy for on-site equipment until such time that normal power to the site can be restored. 

This capability will be provided by 4160 Vac turbine generators provided from the NSRC. Two 

4160 Vac turbine generators will be provided from an NSRC in order to supply power to a 2400 

Vac bus through a plant supplied 4160/2400 V transformer. By restoring the Class 1 E 2400 Vac 

bus, power can be restored to the CACs or SOC. 

The NRC staff reviewed Attachment 9 of Engineering Change 46465 (Phase 3 4160 Vac 

Turbine Generator Sizing), conceptual single line electrical diagrams, the separation and 

isolation of the FLEX turbine generators from the Class 1 E EOGs, and procedures that direct 

operators how to align, connect, and protect associated systems and components. Based on its 

review of the sizing calculation, the NRC staff confirmed that two 1-MW 4160 Vac turbine 
- 37 -

generators will provide sufficient capacity and capability to supply the necessary loads following 

an ELAP to maintain core cooling and containment. Refer to Section 3.2.3.6 above for 

additional analyses. 

3.4.5 Conclusions 

Based on this evaluation, the NRC staff concludes that the licensee has developed guidance 

that, if implemented appropriately, should maintain or restore containment functions following an 

ELAP event consistent with NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01, and should 

adequately address the requirements of the order. 

3.5 Characterization of External Hazards 

Sections 4 through 9 of NEI 12-06, Rev. 0, provide the methodology to identify and characterize 

the applicable BDBEEs for each site. In addition, NEI 12-06 provides a process to identify 

potential complicating factors for the protection and deployment of equipment needed for 

mitigation of site-specific external hazards leading to an ELAP and LUHS. 

Characterization of the applicable hazards for a specific site includes the identification of 

realistic timelines for the hazard, characterization of the functional threats due to the hazard, 

development of a strategy for responding to events with warning, and development of a strategy 

for responding to events without warning. 

The licensee reviewed the plant site against NEI 12-06 and determined that FLEX equipment 

should be protected from the following hazards: seismic; external flooding; tornado high winds; 

snow, ice and extreme cold; and extreme high temperatures. 

References to external hazards within the licensee's mitigating strategies and this SE are 

consistent with the guidance in NEl-12-06 and the related NRC endorsement of NEI 12-06 in 

JLD-ISG-2012-01. Guidance document NEI 12-06 directed licensees to proceed with 

evaluating external hazards based on currently available information. For most licensees, this 

meant that the OIP used the current design-basis information for hazard evaluation. Coincident 

with the issuance of Order EA-12-049, on March 12, 2012, the NRC staff issued a Request for 

Information pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f) 

[Reference 20] (hereafter referred to as the 50.54(f) letter), which requested that licensees 

reevaluate the seismic and flooding hazards at their sites using updated hazard information and 

current regulatory guidance and methodologies. Due to the time needed to reevaluate the 

hazards, and for the NRC to review and approve them, the reevaluated hazards were generally 

not available until after the mitigation strategies had been developed. The NRC staff has 

developed a proposed rule, titled "Mitigation of Beyond-Design-Basis Events [MBDBE]," 

hereafter called the MBDBE rule, which was published for comment in the Federal Register on 

November 13, 2015 [Reference 41 ]. The proposed MBDBE rule would make the intent of 

Orders EA-12-049 and EA-12-051 generically applicable to all present and future power reactor 

licensees, while also requiring that licensees consider the reevaluated hazard information 

developed in response to the 50.54(f) letter. 

The NRC staff requested Commission guidance related to the relationship between the 

reevaluated flooding hazards provided in responses to the requested information and the 

requirements for Order EA-12-049 and related rulemaking to address beyond-design-basis 
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external events (see COMSECY-14-0037), "Integration of Mitigating Strategies for BeyondDesign-Basis External Events and the Reevaluation of Flooding Hazards" [Reference 27]. The 

Commission provided guidance in a SRM to COMSECY-14-0037," dated March 30, 2105 

[Reference 21]. The Commission approved the staff's recommendations that licensees need to 

address the reevaluated flooding hazards within their mitigating strategies for BDBEEs, and that 

licensees may need to address some specific flooding scenarios that could significantly impact 

the power plant site by developing scenario-specific mitigating strategies, possibly including 

unconventional measures, to prevent fuel damage in reactor cores or SFPs. The NRC staff did 

not request that the Commission consider making a requirement for mitigating strategies 

capable of addressing the reevaluated flooding hazards be immediately imposed, and the 

Commission did not require immediate imposition. In a letter to licensees dated September 1, 

2015 [Reference 24], the NRC staff informed the licensees that the implementation of mitigation 

strategies should continue as described in licensee's OIPs, and that the related NRC SEs and 

inspections will rely on the guidance provided in JLD-ISG-2012-01, Rev. 0, and the related 

industry guidance in Rev. 0 to NEI 12-06. The hazard reevaluations may also identify issues to 

be entered into the licensee's corrective action program consistent with the OIPs submitted in 

accordance with Order EA-12-049. 

As discussed above, licensees are reevaluating the site seismic and flood hazards as requested 

in the NRC's 50.54(f) letter. After the NRC staff approves the reevaluated hazards, licensees 

will use this information to perform flood and seismic mitigating strategies assessments (MSAs) 

per the guidance in NEI 12-06, Rev. 2, Appendices G and H [Reference 42]. The NRC staff 

endorsed Rev. 2 of NEI 12-06 in JLD-ISG-2012-01, Rev. 1 [Reference 43]. The licensee's 

MSAs will evaluate the mitigating strategies described in this SE using the revised seismic 

hazard information and, if necessary, make changes to the strategies or equipment. Licensees 

will submit the MSAs for NRC staff review. 

The licensee developed its OIP for mitigation strategies by considering the guidance in NEI 12-

06 and the site's design-basis hazards. Therefore, this SE makes a determination based on the 

licensee's OIP and FIP. The characterization of the applicable external hazards for the plant 

site is discussed below. 

3.5.1 Seismic 

In its FIP, the licensee stated that per the UFSAR, the seismic criteria for PNP includes two 

design-basis earthquake spectra: operating basis earthquake (OBE) and safe shutdown 

earthquake (SSE). A conservative SSE having peak horizontal ground acceleration (g) at the 

surface of 0.05 g is the recommended value in the UFSAR. However, to be conservative a 

hypothetical SSE of 0.2 g has been selected for designs and analyses, per UFSAR Section 

2.4.4. It should be noted that the actual seismic hazard involves a spectral graph of the 

acceleration versus the frequency of the motion. Peak acceleration in a certain frequency 

range, such as the numbers above, is often used as a shortened way to describe the hazard. 

As the licensee's seismic reevaluation activities are completed, the licensee is expected to 

assess the mitigation strategies to ensure they can be implemented under the reevaluated 

hazard conditions as will potentially be required by the proposed MBDBE rulemaking. The 

licensee has appropriately screened in this external hazard and identified the hazard levels to 

be evaluated. 
- 39 -

3.5.2 Flooding 

The licensee stated in its FIP, that the flood assessment for the PNP site provided in the PNP 

UFSAR considered flooding due to high levels in Lake Michigan, high rainfall, wind-generated 

waves concurrent with flooding, and seiches. High tides, hurricane surges, and tsunamis were 

determined to not affect the site due to the inland location. FLEX equipment is required to be 

accessible, deployable and functioning upon an external flood per NEI 12-06. PNP is 

susceptible to seiche and precipitation hazards. Palisades UFSAR Section 5.4.1.1 states that 

the current probable maximum flood (PMF) due to maximum expected precipitation is no higher 

than 6 inches (in.) above grade due to overland run-off to Lake Michigan. General plant grade 

elevation is 589 ft. -0 in. along Lake Michigan. Flooding due to precipitation is considered slow 

moving, with warning time of a few days and unknown duration. 

The FIP also stated that the flood expected from a seiche would reach an elevation of 594.1 ft. 

per UFSAR Section 2.2.2.1. A seiche is defined as a standing wave in a partially enclosed body 

of water (i.e. a lake) and is characterized as an unpredictable event with no warning but is 

expected to recede in 30 minutes. FLEX Phase 2 portable equipment deployment will be 

initiated after the seiche has receded, and FLEX equipment is stored and will be staged above 

the PMF. All plant equipment that will be used for FLEX is protected against a flood to a level of 

594.4 ft. per UFSAR Section 2.2. 

As the licensee's flooding reevaluation activities are completed, the licensee is expected to 

assess the mitigation strategies to ensure they can be implemented under the reevaluated 

hazard conditions as will potentially be required by the proposed MBDBE rulemaking. The 

licensee has appropriately screened in this external hazard and identified the hazard levels to 

be evaluated. 

3.5.3 High Winds 

In NEI 12-06, Section 7, provides the NRG-endorsed screening process for evaluation of high 

wind hazards. This screening process considers the hazard due to hurricanes and tornadoes. 

The screening for high wind hazards associated with hurricanes should be accomplished by 

comparing the site location to NEI 12-06, Figure 7-1 (Figure 3-1 of U.S. NRC, "Technical Basis 

for Regulatory Guidance on Design Basis Hurricane Wind Speeds for Nuclear Power Plants," 

NUREG/CR-7005, December, 2009), if the resulting frequency of recurrence of hurricanes with 

wind speeds in excess of 130 mph exceeds 1 E-6 per year, the site should address hazards due 

to extreme high winds associated with hurricanes using the current licensing basis for 

hurricanes. 

The screening for high wind hazard associated with tornadoes should be accomplished by 

comparing the site location to NEI 12-06, Figure 7-2, from U.S. NRC, "Tornado Climatology of 

the Contiguous United States," NUREG/CR-4461, Rev. 2, February 2007; if the recommended 

tornado design wind speed for a 1 E-6/year probability exceeds 130 mph, the site should 

address hazards due to extreme high winds associated with tornadoes using the current 

licensing basis for tornados or Regulatory Guide 1.76, Rev. 1. 
- 40 -

The licensee stated in its FIP, that SSCs credited for FLEX are evaluated for the design-basis 

tornado for PNP. The tornado design parameters are an extreme design event that bounds 

those parameters defined in the UFSAR for operating basis straight line wind speeds. PNP 

design-basis wind from tornado on existing class 1 SSCs results in an external pressure 

resulting from a tornado funnel having a tangential wind speed of 300 mph and a differential 

pressure of 3 psig. The tornado missile criterion varies for different class 1 SSCs within the 

plant. The NRC reviewed these missiles against NUREG-0800 (UFSAR, Section 5.5.1.3.2). 

The containment structure, auxiliary building, turbine building, intake structure, auxiliary building 

radwaste addition, auxiliary building addition, diesel fuel oil storage tank housing, and 

condensate storage tank tornado missile design parameters are defined in the UFSAR. All 

other SSCs for which there is no existing design-basis have been evaluated per NRC 

Regulatory Guide 1.76, Rev. 1, based on NUREG-0800, Chapter 3. Although the licensee did 

not address the impact of a hurricane in the integrated plan, the site is beyond the range of high 

winds from a hurricane per NEI 12-06 Figure 7-1. 

Therefore, high-wind hazards are applicable to the plant site. The licensee has appropriately 

screened in the high wind hazard and characterized the hazard in terms of wind velocities and 

wind-borne missiles. 

3.5.4 Snow. Ice, and Extreme Cold 

As discussed in NEI 12-06, Section 8.2.1, all sites should consider the temperature ranges and 

weather conditions for their site in storing and deploying FLEX equipment consistent with 

normal design practices. All sites outside of Southern California, Arizona, the Gulf Coast and 

Florida are expected to address deployment for conditions of snow, ice, and extreme cold. All 

sites located north of the 35th Parallel should provide the capability to address extreme snowfall 

with snow removal equipment. Finally, all sites except for those within Level 1 and 2 of the 

maximum ice storm severity map contained in Figure 8-2 should address the impact of ice 

storms. 

In its FIP, regarding the determination of applicable extreme external hazards, the licensee 

stated that the site is located at 42°18' N, - 86°18' W, and 42°20' N, - 86°19' W ... In addition, 

the site is located within the region characterized by the Electric Power Research Institute 

(EPRI) as ice severity level 5 (NEI 12-06, Figure 8-2, Maximum Ice Storm Severity Maps). 

Consequently, the site is subject to severe icing conditions that could cause severe damage to 

electrical transmission lines. The licensee concludes that the plant screens in for an 

assessment for snow, ice, and extreme cold hazard. In its FIP, the licensee stated that FLEX 

equipment is protected from severe temperatures. 

In summary, based on the available local data and Figures 8-1 and 8-2 of NEI 12-06, the plant 

site does experience significant amounts of snow, ice, and extreme cold temperatures; 

therefore, the hazard is screened in. The licensee has appropriately screened in the hazard 

and characterized the hazard in terms of expected temperatures. 

3.5.5 Extreme Heat 

The licensee stated in its FIP, that due to PNPs' proximity to Lake Michigan, the site 

experiences cooler temperatures than most locations in their region, the 10-year maximum 

temperature is 95°F per UFSAR Section 2.5. However, the guidance in NEI 12-06 Section 9.2 
- 41 -

states that all sites within the continental United States will address the high temperature 

scenarios. Thus, the PNP site screens in for the extreme heat hazard. The NEI 12-06 guidance 

requires consideration of extreme high temperatures up to 110-120°F. Per UFSAR Sections 

2.5.1 and 5.9.1.4, because of the proximity of the cool lake, the extreme 10-year maximum 

temperature is 95°F. Deployable FLEX equipment is designed for high temperature operation 

and stored/maintained at a temperature that will ensure it will function when deployed. 

In summary, based on the available local data and the guidance in Section 9 of NEI 12-06, the 

plant site does experience extreme high temperatures. The licensee has appropriately 

screened in the high temperature hazard and characterized the hazard in terms of expected 

temperatures. 

3.5.6 Conclusions 

Based on the evaluation above, the NRC staff concludes that the licensee has developed a 

characterization of external hazards that is consistent with NEI 12-06 guidance, as endorsed by 

JLD-ISG-2012-01 and should adequately address the requirements of the order in regard to the 

characterization of external hazards. 

3.6 Planned Protection of FLEX Equipment 

3.6.1 Protection from External Hazards 

The Palisades FLEX storage configuration utilizes two storage locations; one located inside the 

protected area (PA) and one located outside the PA. FLEX storage building "A" is a preengineered metal building located inside the PA. FLEX storage building "A" protects the stored 

FLEX equipment from extreme temperatures, ice and heavy snow, PMF and seismic BDBEE. 

FLEX storage building "A" does not protect the stored FLEX equipment against a high wind 

BDBEE or tornado missiles. 

FLEX storage building "B" is a hardened reinforced concrete building located outside the PA. 

FLEX storage building "B" protects the stored FLEX equipment from extreme temperatures, ice 

and heavy snow, probable maximum flooding and high wind BDBEE, including tornado missiles. 

FLEX building "B" itself protects against a seismic BDBEE. However, liquefaction concerns due 

to a seismic BDBEE were identified for the equipment haul path areas between FLEX building B 

and the PA. Therefore, FLEX equipment stored in FLEX building "B" cannot be credited for 

successful Phase 2 deployment into the PA staging areas in a seismic BDBEE. However, FLEX 

building "B" will also store selected Phase 3 FLEX equipment. The Phase 3 equipment stored 

in the FLEX storage building "B" is considered accessible in the case of a seismic event as it is 

not required until 72 hours (or later) following the BDBEE, which would allow adequate time to 

resolve accessibility issues due to a seismic BDBEE. 

Guidance document NEI 12-06, Rev. 0, Section 10.1, stipulates that N+ 1 sets of FLEX 

equipment be protected from all applicable BDBEEs. While each FLEX storage building ("A" 

and "B") stores a complete set of FLEX equipment, including debris removal equipment, both 

storage locations are required to ensure that a single set (N) of FLEX equipment would be 

available for all applicable hazards. In its FIP, the licensee acknowledges that the Palisades 

FLEX storage configuration is an alternative to the guidance in NEI 12-06, Rev. 0. The FIP also 
- 42 -

describes, a reduced allowed out of service time for FLEX equipment maintenance to address 

the alternative configuration as discussed in Section 3.14.2 of this SE. Below are additional 

details on how FLEX equipment is protected from each of the external hazards. 

3.6.1.1 Seismic 

As discussed above, both storage buildings "A" and "B" have been shown to have sufficient 

margin to withstand SSE seismic loading. However, due to liquefaction concerns, deployment 

of the FLEX equipment in storage building "B," located outside the PA, may not be possible 

following a seismic event and, therefore, will not be credited for the seismic event. As noted in 

Section 3.14.2 below, the licensee has taken an alternate approach to the guidance of NEI 12-

06, Rev. O, for FLEX equipment storage regarding the seismic hazard utilizing a reduced outage 

time for FLEX equipment maintenance. 

3.6.1.2 Flooding 

As discussed above in Section 3.5.2 of this SE, both storage buildings "A" and "B" are designed 

to be protected against a flooding BDBEE. 

3.6.1.3 High Winds 

As discussed above, only storage building "B" is designed to be protected against high wind 

(tornado hazards). As noted above, the licensee had taken an alternate approach to the 

guidance of NEI 12-06, Rev. 0, for FLEX equipment storage regarding the high wind hazard. 

3.6.1.4 Snow. Ice. Extreme Cold and Extreme Heat 

As discussed above, both storage buildings "A" and "B" are designed to be protected against 

ice, snow, and extreme temperature conditions. The buildings will not be provided with heating 

or cooling capabilities, but other measures such as block heaters and natural ventilation will be 

used to maintain equipment ready for use in extreme temperature conditions. 

3.6.2 Reliability of FLEX Equipment 

Section 3.2.2 of NEI 12-06 states, in part, that in order to assure reliability and availability of the 

FLEX equipment, the site should have sufficient equipment to address all functions at all units 

on-site, plus one additional spare (i.e., an N+1 capability, where "N" is the number of units on 

site). It is also acceptable to have a single resource that is sized to support the required 

functions for multiple units at a site (e.g., a single pump capable of all water supply functions for 

a dual unit site). In this case, the N+1 could simply involve a second pump of equivalent 

capability. In addition, it is also acceptable to have multiple strategies to accomplish a function, 

in which case the equipment associated with each strategy does not require an additional spare. 

While each FLEX storage building ("A" and "B") stores a complete set of FLEX equipment, 

including debris removal equipment, both storage locations are required to ensure that a single 

set (N) of FLEX equipment would be available for all applicable hazards. In its FIP, the licensee 

acknowledges that the Palisdaes FLEX storage configuration is an alternative to the guidance in 
- 43 -

NEI 12-06, Rev. 0. The FIP also describes a reduced allowed out of service time for FLEX 

equipment maintenance to compensate for the alternative FLEX storage configuration as 

discussed in Section 3.14.2 of this SE. 

3.6.3 Conclusions 

Based on this evaluation, the NRC staff concludes that the licensee has developed guidance 

that, if implemented appropriately, should protect the FLEX equipment during a BDBEE 

consistent with NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01, using an alternate 

approach for storage and maintenance of FLEX equipment, and should adequately address the 

requirements of the order. 

3.7 Planned Deployment of FLEX Equipment 

In its FIP, the licensee described the PNP FLEX response strategies plan for deployment of 

pumps, DGs, and other equipment from the FLEX buildings to locations at the power block to 

support the various FLEX capabilities. 

3.7.1 Means of Deployment 

Areas adjacent to the equipment storage facilities and staging areas, as well as the deployment 

and routing paths, are kept normally accessible including removal of ice and heavy snow. FLEX 

building "A," located within the PA, stores a heavy duty pickup truck (with plow) that also 

supports towing of the FLEX equipment. The licensee's debris assessment concluded that the 

debris impact on the relatively short FLEX deployment path within the PA would be minimal for 

the external hazards addressed by FLEX building "A." Therefore, any debris would be expected 

to be removed using the heavy duty pickup truck. 

FLEX building "B," located outside the PA, is credited for mitigation of a high wind event. The 

FLEX deployment path associated FLEX building "B" requires entrance into the PA through 

multiple security fences and barriers. A greater debris impact is expected on the FLEX 

deployment path associated with FLEX building "B" due to the longer deployment route and the 

propensity of the high wind event to generate debris. In addition to the heavy duty pickup truck, 

FLEX building "B" will also store a large Case 821 F wheeled loader to remove any expected 

debris. 

3.7.2 Deployment Strategies 

As described in Section 3.6.1 above, liquefaction concerns due to a seismic BDBEE were 

identified for the equipment haul path areas between FLEX building Band the PA. Therefore, 

FLEX equipment stored in FLEX building "B" cannot be credited for successful Phase 2 

deployment into the PA staging areas in a seismic BDBEE. This is compensated for by the 

alternate approach to maintenance times noted in Section 3.14.2 below. However, FLEX 

building "B" will also store selected Phase 3 FLEX equipment. The Phase 3 equipment stored 

in the FLEX storage building "B" is considered accessible in the case of a seismic event as it is 

not required until 72 hours (or later) following the BDBEE, which would allow adequate time to 

resolve accessibility issues due to a seismic BDBEE. 
- 44 -

The licensee also performed an assessment (ADAMS Accession No. ML 16174A171) regarding 

the potential for frazil ice development at the FLEX suction point of the UHS that determined 

that frazil ice would not impact the deployment/operation of FLEX equipment. This is due to the 

UHS intake structure configuration and the ability to recirculate warmer water from the FLEX 

pump back to the UHS intake inhibiting the formation of frazile ice. 

3.7.3 Connection Points 

3.7.3.1 Mechanical Connection Points 

Core Cooling 

The portable diesel driven FLEX pump will be staged west of the Intake Structure. A 6 in. 

suction hose and strainer will be lowered through a removed section of grating along the west 

wall of the intake structure and will be routed from the discharge of the FLEX pump to the hose 

distribution manifold located in the Turbine Building. The hose route paths pass through the 

turbine building to utilize the connection points in the MFW or AFW system where a new 

connection point is installed for this purpose. During the audit, the licensee stated that there is 

enough hose to route from the FLEX pump outside the Turbine Building, directly into the 

Auxiliary Building. The staff's evaluation of the robustness and availability of the UHS for an 

ELAP event is discussed in Section 3.10.3. 

In the FIP, Section 2.3.4.16 states that the primary connection point is located in the MFW 

system between valves MV-FW907 and MV-FW909, on Feedwater Heater E-6B. The 

connection is a pipe tee on the existing MFW line that contains a hose connection and two new 

isolation valves that are normally closed. EC-46465, "FLEX Basis," Rev. 0, Section 3.1.2.2, 

evaluates the credited portion of the MFW system as robust for all applicable hazards. 

Additionally, Fl P Section 2.3.4.17 states that the alternate connection is located in the AFW 

system and is installed in the discharge piping of the motor-driven AFW pump P-8C. 

Specifically, a new pipe branch was added between valves CV-736A and CV-737A, and the 

pipe branch includes two, normally shut isolation valves. In the Palisades UFSAR, Table 5.2-3, 

indicates that the AFW piping is Design Class 1 with this connection point located in the West 

Safeguards Room of the Auxiliary Building; thus, this connection point is protected from all 

applicable external hazards. EC-46465, "FLEX Basis," the AFW connection is designed 

consistent with Design Class 1 standards as described in UFSAR, Rev. 29, Section 5.10.1.1. 

Since the connections are in two diverse systems, the staff finds the licensee is consistent with 

NEI 12-06, Table D-1, and the licensee has the capability of SG make-up with a portable pump 

through primary and alternate injection points to inject through separate divisions/trains. The 

staff also noted that the primary and alternate FLEX pump discharge connection points and 

injection paths should be available during an ELAP event consistent with NEI 12-06, Section 

3.2.2. 

The portable electric driven FLEX air compressor that supports long-term ADV operation is 

powered by the portable diesel driven FLEX generator and staged within the Auxiliary Building. 

An extension cable will be run from either the primary or secondary generator staging location to 

the compressor staging location. As stated in Section 3.2.3.1.1, Nitrogen Backup Station 9 is 
- 45 -

located in the Auxiliary Building; thus, the air hose connection located on that station is 

protected from all applicable external hazards and should be available during an ELAP event 

consistent with NEI 12-06, Section 3.2.2. 

The licensee's initial Phase 3 strategy relies on continuing Phase 2 strategies with additional 

NSRC equipment. However, the licensee will place the PCS on SOC. The SWS flow to cool 

the CCW heat exchangers is provided by removing the cover of the SWS pump strainer F-2C, 

and connecting a deployable hose connection header consisting of six 5 in. STORZ hose 

connections and a 16 in. flanged connection to allow for tie into the SWS piping. As stated in 

the FIP, the SWS connection piping is located within the Design Class 1 Intake Structure, and 

the SWS piping is Design Class 1; thus, the connection point is protected from all applicable 

external hazards and should be available during an ELAP event consistent with NEI 12-06, 

Section 3.2.2. 

PCS Inventory Control 

The Palisades FLEX strategy for Modes 1 through 4 does not require connections for portable 

FLEX equipment to the PCS, as the re-powered charging pumps are used to supply borated 

water from the BAST and the boric acid mixing tank T-77 to the PCS. However, connection 

points are required for the FLEX strategy for Modes 5 and 6. Newly installed connections are 

provided in the HPSI and LPSI systems to support direct injection to the PCS in Mode 5 or 6 or 

during Phase 3 as well. 

SFP Cooling 

There are no permanent system piping connections associated with the SFP makeup via hose 

directly to the pool or spray using portable monitor nozzles. All equipment is portable and does 

not require any physical connections to permanent plant equipment. For SFP makeup using 

installed SFP cooling piping, an existing hose connection and isolation valve located in the SFP 

heat exchanger room are used and do not require access to the SFP floor. As described in the 

FIP, the sections of the SFP cooling system relied on are designated Design Class 1 and are 

located in the Auxiliary Building, which is robust with respect to all applicable external hazards. 

3.7.3.2 Electrical Connection Points 

Electrical connection points are only applicable for Phases 2 and 3 of the licensee's mitigation 

strategies for a BDBEE. During Phase 2, the licensee has developed a primary and alternate 

strategy for supplying power to equipment required to maintain or restore core cooling, 

containment, and SFP cooling using a combination of permanently installed and portable 

components. There are two portable FLEX 480 Vac diesel generators provided for the strategy, 

one is stored in FLEX Building "A" and one in FLEX Building "B." The licensee will deploy a 

single 480 Vac 463 kW portable diesel generator during an ELAP event. The primary staging 

location for the FLEX diesel generator is northeast of the Containment Building. The licensee 

will route cables from the staging location to quick disconnect box EP-1901 in the North 

Penetration Area. The secondary staging location for the FLEX diesel generator is the 

courtyard north of the Turbine Building. The licensee will route cables from the staging location 

to the Electrical Equipment Room (EER) and over to quick disconnect box EP-2001. 
- 46 -

Primary and alternate connections for the portable FLEX diesel generator are accomplished 

utilizing breakers installed in previously spare cubicles of 480 Vac LCCs 19 and 20. The 

primary connection is at LCC-19 and the alternate is LCC-20. The connection to LCC-19 will be 

via a new FLEX panel (EP-1901) installed in the North Penetration Area with one set of quick 

disconnect receptacles. Permanent cables complete the connection from EP-1901 to breaker 

52-19FLEX in LCC-19. Energizing LCC-19 provides the means to power up MCCs 1, 21, and 

23, and backfeed LCC-11. The LCC-11 can be cross-tied to LCC-12 through an existing 

permanently installed Class 1 E breaker. The connection to LCC-20 will be via a new FLEX 

panel (EP-2001) installed in the EER with one set of quick disconnect receptacles. Permanent 

cables complete the connection from EP-2001 to breaker 52-20FLEX in LCC-20. Energizing 

LCC-20 provides the means to power up MCCs 2, 22 and 24, and backfeed LCC-12, which can 

be cross-tied to LCC-11 through an existing permanently installed Class 1 E breaker. 

Phase rotation for the 480 Vac diesel generators have been confirmed and will be controlled 

using color coded cables. 

The MCCs provide power to various system MO Vs required by the strategy, Class 1 E Battery 

Chargers, emergency lighting panels, Charging Pumps and other components used in the 

Palisades FLEX strategy. The LCCs, MCCs and breakers are all safety-related, Class 1 E 

components located within the Seismic Category 1 Auxiliary Building. As such, these SSCs are 

fully protected and qualified for all applicable external events. The downstream battery chargers 

and Class 1 E loads are seismically robust, protected within the Auxiliary Building and are fully 

qualified for all applicable external events. 

For Phase 3, the NSRC will supply two 4160 Vac, 1-MW combustion turbine generators to 

provide a combined capacity of 2 MW. The 4160 Vac NSRC turbine generators will be 

connected to one train of Class 1 E 2400 Vac switchgear via the Step Down Transformer to 

either Bus 1 C or Bus 1 D. The transformer will step the voltage from the NSRC generator (4160 

Vac) down to the Bus 1 C (1 D) voltage (2400 Vac). The 4160 Vac/2400 Vac Step Down 

Transformer will be connected to Bus 1 D or 1 C using temporary cables connected directly to an 

existing breaker 152-213 on Bus 1 D (or 152-107 on Bus 1 C) on the bus for all 3 phases. The 

switchgear is located within the Seismic Category 1 Auxiliary Building. All associated 

downstream switchgear providing power to loads used in the Phase 3 strategy are also safetyrelated, Class 1 E SSCs and are also located within the Auxiliary Building. All electrical circuits 

are fully protected and qualified for all applicable external events. No modification or designed 

connection point is required for establishing the connection to the switchgear. The actions 

necessary to connect the cables are contained in FLEX procedures and plant staff is trained to 

perform the activity. Palisades stores the Phase 3, 4160Vac/2400Vac, Step Down Transformer 

in FLEX Building "B", which provides protection from all applicable external hazards. 

Phase rotation for the 4160 Vac turbine generators will be confirmed and controlled using color 

coded cables and phase rotation meters in the transformer cabinet. 

3.7.4 Accessibility and Lighting 

In its FIP, the licensee stated that the MGR is served by emergency ac lighting, emergency de 

lighting (powered by station batteries) and battery powered emergency lighting units (ELUs). 

During Phase 1 of a BDBEE, lighting in the MGR will be provided by the emergency de lighting 

and ELUs. At 8 hours after the event (start of Phase 2) the portable diesel driven FLEX 
- 47 -

generator is connected to provide power to the battery chargers for the continued operation of 

emergency de lighting. Operators will have portable LED flashlights and lanterns available as 

necessary to provide lighting for operator actions outside the MCR. Area lighting is required for 

outside deployment during the night and it is met by the lights on the truck used to haul the 

generator, as well as portable diesel driven FLEX light towers. Additional battery operated LED 

light towers are also stored in each FLEX storage building. 

3.7.5 Access to Protected and Vital Areas 

In its FIP, the licensee stated that the pathway from FLEX storage building "B" requires 

navigation through multiple security barriers and requires assistance from security personnel. 

The control of security gates after a FLEX event will be through the security program. 

3.7.6 Fueling of FLEX Equipment 

The FIP states that fueling of FLEX equipment is addressed in an evaluation that considered the 

amount of diesel fuel required to operate the portable diesel driven FLEX generator and one 

portable diesel driven FLEX pump from hours 8 to 72 (64 hours) following a BDBEE. 

Additionally, the evaluation identified an acceptable source of diesel fuel on site, the diesel fuel 

oil storage tank (DFOST) (T-10A), and determined the pumping requirements to ensure that this 

fuel is accessible during an ELAP. During the audit, the licensee stated the amount of diesel 

fuel consumed by the FLEX portable pump and the generator is approximately 3,300 gallons for 

a period of 64 hours. As stated in the FIP, the minimum required storage volume for tank T-10A 

is 30,554 gallons. In the UFSAR Table 3.2-1 indicates that the DFOST T-10A is a Design Class 

1 component and protected from all applicable hazards. Therefore, the diesel fuel stored in tank 

T-10A is more than sufficient to provide the fuel oil required by the FLEX equipment to operate 

for approximately 24 days and provides sufficient time for the licensee to obtain additional fuel 

oil from off-site. 

The strategy for refueling the portable diesel driven FLEX pump and portable diesel driven 

FLEX generator is to use a portable, de powered fuel pump to transfer diesel fuel from tank 

T-10A into a 500 gallon tank mounted trailer. This de fuel pump will be powered from the heavy 

duty FLEX truck wiring harness. The trailer towed by the heavy duty FLEX truck will be used to 

deliver fuel to the portable diesel driven FLEX pump and diesel generator, stopping by tank T1 OA to refill the trailer as needed. A second de powered fuel pump, powered from the FLEX 

truck, will be mounted on the trailer, and will be used to transfer fuel from the trailer to each 

piece of equipment. The evaluation estimates that the initial transfer of fuel from tank T-1 OA to 

each piece of FLEX equipment will require approximately 2 hours. Once the initial transfer of 

fuel from tank T-10A is complete, fuel can be re-delivered to either set of equipment in intervals 

of 90 minutes. The limiting FLEX equipment with regard to refuel frequency is the FLEX 

portable pump, which is conservatively estimated to require a refuel every 5 hours. 

Two sets of FLEX refueling equipment (trailer mounted tank, pumps, etc.) are provided, with 

one set being stored in FLEX storage building "A" and the other in FLEX storage building "B". 

The fuel oil transfer tank is stored full of fuel (greater than 95 percent capacity). Each FLEX 

pump and generator is stored with fuel in their tanks. Fuel in each generator and pump is 

sampled and replaced on a preventive maintenance interval specified by Entergy fleet PM basis 

templates specific to the equipment. Diesel fuel in the tank T-10A is low sulfur (500 ppm sulfur) 

fuel oil or better, and is routinely sampled and tested commensurate with its use for the onsite 
- 48 -

EDGs to assure fuel oil quality is maintained to American Society for Testing Material standards, 

which ensures this fuel oil is compatible with the Environmental Protection Agency tier 2 and tier 

3 engines (which will run on low or ultra-low sulfur fuel) for the portable FLEX pumps and 

generators. 

During the onsite audit, the licensee stated that the quality of fuel stored in the FLEX equipment 

will be confirmed through a periodic testing program. Additionally, the staff walked down the 

refueling strategy with licensee representatives. Palisades procedure FIG-10, "Fuel Oil 

Transfer," Rev. 0, contains refueling instructions for the FLEX equipment. This procedure also 

contains a table with fill times and consumption rates for the FLEX equipment. Based on the 

guidance provided in this procedure, the staff finds it is reasonable that diesel-powered FLEX 

equipment should be fueled and refueled to ensure uninterrupted operation to support the 

licensee's FLEX strategies. 

3.7.7 Conclusions 

The NRC staff concludes that the licensee has developed guidance that, if implemented 

appropriately, should allow deploying the FLEX equipment following a BDBEE consistent with 

NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01, using an alternate approach for 

maintenance of FLEX equipment, and should adequately address the requirements of the order. 

3.8 Considerations in Using Offsite Resources 

3.8.1 PNP SAFER Plan 

The licensee stated in its FIP, that the industry established two NSRCs that house backup 

equipment to be used by nuclear plant sites and additional equipment for long-term recovery in 

support of utilities during BDBEEs. One facility is located in Phoenix, Arizona and the other is in 

Memphis, Tennessee. Each NSRC holds five sets of equipment, four of which will be able to be 

fully deployed to the plant when requested. The fifth set allows removal of equipment from 

availability to conduct maintenance cycles. In addition, the plant's FLEX equipment hose and 

cable end fittings are standardized with the equipment supplied from the NSRC. Entergy has 

established contracts with the pooled equipment inventory company (PEICo) that operates 

these facilities and the NSRCs are obligated to provide one complete set of the sites Phase 2 

equipment, as well as requested long-term recovery equipment unique to the sites strategy. 

The PNP on-site BDBEE equipment, hose and cable end fittings are standardized with the 

equipment supplied from the NSRC. In the event of a BDBEE and subsequent ELAP/LUHS 

condition, equipment will be moved from the NSRC to a local assembly area established by the 

SAFER team. The NSRC equipment will begin arriving at PNP designated site staging 

locations within the first 24 hours and all obligated equipment is received, staged and deployed 

no later than 72 hours after the event. 

The equipment stored and maintained at the NSRC for transportation to the local assembly area 

to support the response to a BDBEE at PNP is listed in the FIP. Table 3 identifies the 

equipment that is specifically credited in the FLEX strategies for PNP. 

By letter dated September 26, 2014 [Reference 22], the NRC staff issued its staff assessment of 

the NSRCs established in response to Order EA-12-049. In its assessment, the staff concluded 

that SAFER has procured equipment, implemented appropriate processes to maintain the 
- 49 -

equipment, and developed plans to deliver the equipment needed to support site responses to 

BDBEEs, consistent with NEI 12-06 guidance; therefore, the staff concluded in its assessment 

that licensees can reference the SAFER program and implement their SAFER Response Plans 

to meet the Phase 3 requirements of Order EA-12-049. 

3.8.2 Staging Areas 

In general, up to four staging areas for NSRC supplied Phase 3 equipment are identified in the 

SAFER Plans for each reactor site. These are a Primary (Area C) and an Alternate (Area D), if 

available, which are offsite areas (within about 25 miles of the plant) utilized for receipt of 

ground transported or airlifted equipment from the NSRCs. From Staging Areas C and/or D, the 

SAFER team will transport the Phase 3 equipment to the on-site Staging Area B for interim 

staging prior to it being transported to the final location in the plant (Staging Area A) for use in 

Phase 3. For PNP, the Alternate Staging Area Dis Andrews University Airpark located in 

Berrien Springs, Ml. Staging Area C is the West Michigan Regional Airport, located in Holland, 

Ml. Staging Area B is PNP training building parking lot. Staging Area A is comprised of several 

primary and alternate locations around the plant where the FLEX equipment will be deployed. 

Use of helicopters to transport equipment from Staging Area C to Staging Area B is recognized 

as a potential need within the PNP SAFER Plan and is provided for. 

3.8.3 Conclusions 

Based on this evaluation, the NRG staff concludes that the licensee has developed guidance 

that, if implemented appropriately, should allow utilization of offsite resources following a 

BDBEE consistent with NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01 and adequately 

addresses the requirements of the order. 

3.9 Habitability and Operations 

3.9.1 Equipment Operating Conditions 

3.9.1.1 Loss of Ventilation and Cooling 

Following a BDBEE and subsequent ELAP event at Palisades, ventilation that provides cooling 

to occupied areas and areas containing required equipment will be lost. Per the guidance given 

in NEI 12-06, FLEX strategies must be capable of execution under the adverse conditions 

(unavailability of installed plant lighting, ventilation, etc.) expected following a BDBEE resulting 

in an ELAP/LUHS. The key areas identified for all phases of execution of the FLEX strategy 

activities are the MGR, TDAFW Pump Room, ADV Room, Charging Pump Rooms, Cable 

Spreading Room (CSR), and EER, Battery Rooms, Containment, and areas containing ADVs 

and ADV controls. The licensee has evaluated these areas to determine the temperature 

profiles following an ELAP/LUHS event. 

The primary concern with regard to ventilation is the heat buildup which occurs with the loss of 

forced ventilation in areas that continue to have heat loads. The licensee performed several 

Loss of ventilation analyses to quantify the maximum steady state temperatures expected in 

specific areas related to FLEX implementation to ensure the environmental conditions remain 

acceptable for personnel habitability and within equipment qualification limits. 
- 50 -

Main Control Room 

During the audit process, the NRC staff reviewed calculation EA-EC-46465-03, "Control Room 

Heatup for Extended Loss of AC Power,'' Rev. 0, which modeled the MCR temperature transient 

when all MCR cooling is lost due to an ELAP event. The calculation uses the GOTHIC 

(Generation of Thermal-Hydraulic Information for Containments) Version 7.2b computer 

program. The calculation showed that opening doors and providing a 4000 cubic feet per 

minute (cfm) portable exhaust fan at or before 75 hours after the event initiation will maintain the 

MCR at a peak value of 104°F. Palisades procedure FIG-8, "Alternate Ventilation,'' Rev. 0, 

directs operators to deploy the temporary fan, which is powered from the portable FLEX 

generator. Based on MCR temperatures remaining below 120°F (the temperature limit, as 

identified in NUMARC-87-00, "Guidelines and Technical Bases for NUMARC Initiatives 

Addressing Station Blackout at Light Water Reactors," Rev. 1, for electronic equipment to be 

able to survive indefinitely), the NRC staff finds that the equipment in the MCR should not be 

adversely impacted by the loss of ventilation as a result of an ELAP event. 

TDAFW Pump Room 

The licensee's Phase 1 core cooling FLEX strategy relies on the TDAFW pump as the motive 

force for providing cooling water to the SGs. During the audit, the licensee provided the staff 

with EA-GOTHIC-AFW-01, "Auxiliary Feedwater Pump Room Heat Up Analysis," Rev. 0. This 

analysis accounted for the maximum design basis ambient conditions and assumed leakage 

from a steam trap in the room. The assumptions used in the analysis include: loss of offsite 

power, no active room ventilation, the turbine building temperature is 110°F, initial AFW room 

temperature is 104 °F, and no credit for heat sinks from the insulated AFW piping in the room. 

The only credited cooling was via natural circulation through the ceiling vent pipe. The 

calculation showed the maximum temperature of the AFW room reached 159°F after 5 days 

with no ventilation required (i.e. open doors, portable fans). The licensee indicated that the 

TDAFW pump can withstand temperatures of up to 160°F without detrimental effects. 

Additionally, the licensee relies on the TDAFW pump for the first 8 hours after event and post 8 

hours, the portable diesel driven FLEX pump will provide makeup water to the SGs using the 

UHS as a water source. Therefore, the staff finds it reasonable that the TDAFW pump should 

remain available during an ELAP event for the timeframe required by the licensee's strategy 

with loss of normal ventilation 

Containment 

With regard to containment instrumentation, the NRC staff reviewed calculation EA-EC46465-

02, "PLP MAAP4 Containment Analysis for BDBEE," Rev. 0. This analysis identifies the 

temperature and pressure inside containment following a BDBEE. The licensee's analysis 

identified the reactor cavity as being the compartment in containment with the highest 

temperature. Based on the results of this analysis, the licensee evaluated the instrumentation 

relied upon during the FLEX strategy to ensure functionality of the equipment following an 

ELAP. The instrumentation evaluated included Steam Generator Pressure and Level, 

Pressurizer Pressure and Level, Hot Leg and Cold Leg Temperature, Nuclear Instrumentation, 

Reactor Vessel Level Instrumentation (RVLMS), Core Exit Thermocouples, Softline Cables 

inside Containment, and Containment Pressure. 
- 51 -

According to the licensee's calculation, Containment temperatures will remain below acceptable 

limits (140°F) for critical instrumentation inside Containment for at least 6 days. Because of this, 

the licensee would rely on equipment received from an NSRC to reduce Containment 

temperature. The equipment from the NSRC that is necessary to support Containment cooling 

includes a low pressure/high flow pump and 4160 Vac turbine generators. The licensee plans 

to deploy this equipment within 72 hours after event initiation. The low pressure/high flow pump 

allows for the restoration of service water flow to equipment such as the CCW system and LPSI 

pump. The 4160 Vac turbine generators allow for the restoration of one complete Class 1 E 

distribution bus (1 D or 1 C) which provide power to a LPSI pump, a CCW pump, SDC flowpath 

valves and 3 CACs. The licensee would use the 4160 Vac turbine generators to start and stop 

CACs as required to maintain temperatures within limits to preserve critical instrumentation. 

Additionally, since the 4160 Vac turbine generators can be used to power one complete Class 

1 E 2400 Vac train, SDC can be placed into service to further reduce primary coolant system 

temperature and pressure, minimizing released energy into the Containment due to leakage. 

Based on the NRG staff's review, instrumentation relied upon for the FLEX strategy should be 

available following an ELAP for continued plant coping. 

ADV Room 

The ADVs are located on the third floor of the CCW room, which is a large, open, multi-story 

room. Under extreme high temperatures, the licensee's procedures (FIG-8) direct operators to 

open one door on the lower level (590' elevation) and one large door on the upper level (625' 

elevation) to the atmosphere to provide chimney effect ventilation for this area. With this natural 

ventilation, temperatures in the area of the ADV would not adversely impact operation of the 

valves as a result of an ELAP event. The ADV local controls are located in the same room as 

the new Nitrogen Station 9, which provides backup nitrogen supply for ADV operation. 

Following depletion of Nitrogen Station 9 (approximately 9 hours into the event), the licensee 

will open the door to this room to route the air compressor hose into the connection location in 

this room. This door on the Auxiliary Building roof opens to atmosphere and provides a 

ventilation path for the room. With this ventilation path established, the staff noted that 

temperatures in the room are not expected to be significantly higher than the outside ambient 

temperature since there is no heat load in this room, and would therefore not adversely impact 

the ADV controls as a result of an ELAP event. 

Charging Pump Rooms 

During the onsite audit, the licensee provided engineering analysis EA-JEF-97-01, "Operation of 

Charging Pumps P-55B & C without CCW flow to the oil coolers," Rev. 0. This evaluation 

determined the equilibrium temperature for the charging pumps while cooling water was isolated 

for 72 hours. This analysis found that the charging pumps could reach a maximum temperature 

of 124.6°F in 72 hours, which is below the maximum temperature of the lube oil of 155°F. 

Additionally, as stated in UFSAR, Rev. 29, Section 9.8.5.2., item 3, the only heat input to the 

charging rooms would be the pumps. The UFSAR states that the room would reach 120°F 

approximately 83 hours after the event and the staff noted lube oil cooling could be established 

using NSRC equipment sometime after 72 hours. The staff finds it reasonable that the charging 

pumps should remain available during an ELAP event with loss of normal ventilation. 
- 52 -

Electrical Equipment Room and Cable Spreading Room 

The EER and CSR are the locations of the electrical connection boxes for the FLEX generator 

and contain the MCCs and LCCs to be repowered. During the onsite audit, the licensee 

provided Calculation EA-EC-46465-04, "BDBEE Temperature Calculation for CSR, EER, and 

Battery Rooms," Rev. O. The focus of this calculation was to determine the expected 

temperature in the EERs, Battery Rooms, and Switchgear Rooms following an ELAP. This 

calculation did not take credit for any doors being opened, assumed a conservative heat load for 

each of these rooms, and assumed a constant outside temperature (i.e., no diurnal variation in 

temperature assumed). The calculation was revised (Rev. 1) to extend the analysis duration to 

120 hours or greater, and to take credit for opening CSR and EER doors 24 hours into the 

event. The operator actions to open doors in the CSR and EER are contained in the licensee's 

FSGs. Palisades procedure FIG-8, "Alternate Ventilation," Rev. 0, contains steps to open both 

doors to the CSR and EER as part of providing ventilation for the Battery Rooms. The FIG 

contains a precaution to establish Battery Room ventilation within 9.5 hours of the event so that 

the doors will be open prior to 9.5 hours after the event. 

Based temperatures remaining at or below 120°F (the temperature limit, as identified in 

NUMARC-87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station 

Blackout at Light Water Reactors," Rev. 1, for electronic equipment to be able to survive 

indefinitely), the NRC staff finds that the equipment in the CSR and EER should not be 

adversely impacted by the loss of ventilation as a result of an ELAP event. 

Battery Rooms 

The NRC staff reviewed calculation EA-EC46465-04, "BDBEE Temperature Calculation for 

CSR, EER, and Battery Rooms," Rev. 1. The focus of this calculation was to determine the 

expected temperature in the EERs, Battery Rooms, and Switchgear Rooms following an ELAP. 

The calculation showed that with extreme high temperatures outside, operator actions to open 

doors may be required to maintain the temperature in these rooms less than 120°F and ensure 

long-term functionality of equipment in these rooms. The operator actions to open doors in the 

CSR and EERs are contained in the licensee's FSGs. 

The Palisades safety-related batteries were manufactured by C&D Technologies. The 

qualification testing performed by C&D Technologies demonstrated the ability to perform under 

elevated operating temperature environments. The testing results indicate that the battery cells 

will perform as required in excess of 200 days under an estimated 122°F. 

The elevated temperature also has an impact by increasing the charging current required to 

maintain the float charging voltage set by the charger. The elevated charging current will in turn 

increase cell water loss through an increase in gassing. Based on this, periodic water addition 

may be required or the float charging voltage reduced per the guidance contained in the C&D 

Technologies vendor manual. If battery cell plate uncovery were to occur, failure issues 

associated with plates being exposed would involve the potential development of sulfation and a 

subsequent reduction in capacity. If loss or failure of a battery string were to occur, the battery 

charger has the capability to carry the anticipated loads indefinitely provided that ac power 

remains available to power the charger. 
- 53 -

Based on its review of the essential station equipment required to support the FLEX mitigation 

strategy, which are primarily located in the Containment, MCR, TDAFW Pump Room, Charging 

Pump Rooms, CSR, EER, Battery Rooms, and areas containing ADVs and ADV controls, the 

NRC staff finds that the equipment should perform their required functions at the expected 

temperatures as a result of loss of ventilation during an ELAP/LUHS event. 

3.9.1.2 Loss of Heating 

In NEI 12-06, Section 3.2.2, guideline (12) states that heat tracing is used at some plants to 

ensure cold weather conditions do not result in freezing important piping and instrumentation 

systems with small diameter piping. In the Palisades FIP, the licensee states that no outside 

staged equipment is credited for FLEX that would need heat tracing to prevent freezing. 

Operation of the TDAFW pump involves steam flow through the turbine and associated piping. 

The TDAFW pump is located in a temperature-controlled area of the turbine building and is 

relied upon immediately at the start of an ELAP event and used for 8 hours until the CST and T81 are exhausted. The staff finds it reasonable that low outside temperatures should not have 

an adverse effect on the TDAFW pump because of its location in an initially temperaturecontrolled area, steam flow through the components provides a heat load, and the components 

are in use early in the ELAP event. 

During the audit, the licensee also stated that the FLEX equipment operated outdoors is 

designed to operate over the outdoor temperature range of -10°F to 95°F, which are the site 

temperature extremes per the Palisades UFSAR Section 9.8.2.1 and Table 9-13, Rev. 29. The 

licensee stated during the audit process that both the FLEX storage building inside the PA and 

the Flex storage Building outside the PA, are protected against extreme temperatures, but 

neither will have their own heating /cooling capabilities. The licensee will utilize block or jacket 

water heaters during cold conditions and open the storage building doors and ventilation louvers 

during hot weather conditions. Furthermore, FSG-5, "Initial Assessment and FLEX Equipment 

Staging," Rev. 0, has steps added to it to prevent freezing of the FLEX equipment such as 

intermittently establish flow, drain, etc. Also, FIG-2, "FLEX Pump Staging and Operation," Rev. 

O, directs operators to establish a recirculation line to help prevent hoses from freezing. 

The licensee analyzed electrolyte temperature for the batteries for an extreme low temperature 

event in Calculation EA-EC46465-04. The calculation showed that the electrolyte temperature 

for the batteries in both Battery Rooms will remain above 70°F (Technical Specification 

operability limit) for the duration that they are credited in the licensee's mitigating strategies (8 

hours). After 8 hours, the FLEX diesel generator will be connected to the Battery Chargers. 

Therefore, the batteries will remain functional during the extreme low temperature event without 

any operator action to provide supplemental heat to the rooms. 

Based on its review of the licensee's Battery Room assessment, the NRC staff finds that the 

Palisades safety-related batteries should perform their required functions at the expected 

temperatures as a result of loss of heating during an ELAP event. 
- 54 -

3.9.1.3 Hydrogen Gas Control in Vital Battery Rooms 

To mitigate hydrogen concentration in the Battery Rooms during battery charging, a portable 

ventilation fan is required. Calculation EA-DBD-1.07-001, "Battery Room Ventilation 

Requirements for Hydrogen Control," Rev. 1, provides the basis for Battery Room hydrogen 

control. The calculation states that when the batteries are being equalized (charged) and 

ventilation is lost, the time to reach two percent hydrogen in the rooms is 1.8 hours. Therefore, 

to prevent hydrogen accumulation above two percent, a portable ventilation/exhaust to the 

Battery Rooms will be deployed by 1.5 hours after battery charging begins, or 9.5 hours after 

initiation of the event. The installed Battery Room ventilation system is not credited because it 

is not seismically qualified and is not protected from wind-generated missiles. The portable 

ventilation fan, which is powered from a 120 volt ac receptacle on the FLEX diesel generator, is 

specified with a minimum exhaust flow of 1,000 cfm. This flow exceeds the minimum flow of 

150 cubic ft. per hour calculated in EA-DBD-1.07-001. Calculation EA-DBD-1.07-001 identified 

the hydrogen concentration in the Battery Room as being 0.005 percent with an exhaust flow of 

1,000 cfm, which is well below the explosive limit for hydrogen (4 percent). Flexible duct that 

connects to the portable fan will be routed from the cable spreading room outside of the turbine 

building such that the hydrogen can be exhausted to the atmosphere, preventing hydrogen 

accumulation in other areas of the plant. 

Based on its review, the NRC staff concludes that hydrogen accumulation in the safety-related 

Battery Rooms should not reach the combustibility limit for hydrogen (4 percent) during an 

ELAP event. 

3.9.2 Personnel Habitability 

3.9.2.1 Main Control Room 

During the audit, the licensee provided a copy of EA-EC-46465-03, "Control Room Heatup for 

Extended Loss of AC Power," Rev. 0. This calculation concluded that, when the mitigating 

actions of opening specified doors in conjunction with starting an exhaust fan at 75 hours, the 

MCR temperature peaks at a temperature of 104°F. The licensee's procedure, FIG-8, 

"Alternate Ventilation," Rev. O, directs operators to take the appropriate actions corresponding 

with the analysis assumptions. 

3.9.2.2 Spent Fuel Pool Area 

See Section 3.3.4.1.1 above for the detailed discussion of ventilation and habitability 

considerations in the SFP Area. In general, the licensee plans to complete hose deployment 

and spray nozzle setup on the SFP refueling floor before conditions degrade to the point that 

boiling of the SFP affects habitability. The licensee also has the ability to add water to the SFP 

from the installed SFP cooling piping without accessing the refueling floor. 

3.9.2.3 Other Plant Areas 

The licensee's Phase1 core cooling FLEX strategies rely on the TDAFW pumps as the motive 

force for providing water to the SGs. The TDAFW pump trip and throttle valve, electronic 

governor, and all associated valves are powered by battery-backed Class 1 E power and 

Nitrogen Backup Stations 1 and 2. However, in the unlikely event that that control power and air 
- 55 -

is lost, the steam supply valve, located in the turbine building above the TDAFW pump room, 

can be manually throttled. In its FIP, the licensee stated that access to the TDAFW pump room 

is not expected to be required, but the TDAFW pump is operated less than 8 hours after the 

event so access would be early in the event. The NRC staff noted that the TDAFW pump is 

located in a temperature-controlled area of the Turbine Building. The staff finds it reasonable 

that the TDAFW pump room will remain accessible in the event access is required. 

In its FIP, the licensee stated the operator actions in the CSR and EER includes routing cables, 

positioning exhaust fans and opening doors, which occurs early in the event at a time when the 

temperature would not have increased significantly. Guideline FIG-1, "FLEX Generator Staging 

and Operation," Rev. 0, directs operators to deploy and connect the generator within 8 hours of 

the start of the event. Furthermore, FIG-8, "Alternate Ventilation," Rev. 0, directs operators to 

stage Battery Room ventilation and open the applicable doors within 9.5 hours of the event. 

Given the licensee's timeline and implementing procedures, the staff finds it reasonable that the 

CSR and EER rooms should remain accessible in the event access is required. 

3.9.3 Conclusions 

The NRC staff concludes that the licensee has developed guidance that, if implemented 

appropriately, should maintain or restore equipment and personnel habitability conditions 

following a BDBEE consistent with NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01, and 

should adequately address the requirements of the order. 

3.10 Water Sources 

3.10.1 Steam Generator Make-Up 

Phase 1 

In the FIP, Section 2.3.1 states that the CST initially provides the main source of water for plant 

cool-down at the initial onset of the ELAP. Within 1 hour, operators will open two normally 

closed valves (CV-2008 and CV-2010) to cross-connect the CST and tank T-81 which will 

provide 8 hours of inventory. 

Palisades UFSAR, Rev. 29, Table 5.2-3 states the CST is a Design Class 1 tank. Palisades 

UFSAR, Rev. 27, Section 9.7.4 states the minimum required level in the tank is 94,280 gallons. 

However, the CST was not fully missile protected; therefore, in EC-48187, "CST Analysis and 

Missile Barrier,'' the licensee designed and installed a CST missile barrier to fully protect the 

tank from all applicable tornado missile hazards. The licensee also evaluated the CST in PLPRPT-13-00050, "Turbine Driven Auxiliary Feedwater (TDAFW) Upgrade and Evaluation for 

FLEX," Rev. O, for all other applicable external hazards and determined it to be robust for all 

external hazards. 

As stated in the FIP, tank T-81 was evaluated in PLP-RPT-13-00050, "Turbine Driven Auxiliary 

Feedwater (TDAFW) Upgrade and Evaluation for FLEX," Rev. 0, and determined the tank is 

robust for all applicable hazards, except for the large missile hazard (e.g., automobile), with the 

following modifications: 
- 56 -

• Installing tornado missile protection of the cross-connect valves (CV-2008 and CV2010). 

• Adding of isolation valves on three sections of piping to maintain T-81 and suction piping 

robust. 

• Tornado missile protecting the T-81 tank. 

Calculation EC-48187, "CST Analysis and Missile Barrier," designed and installed the missile 

protection for the tank and the cross-connect valves; however, since tank T-81 tank is not robust 

with respect to all tornado missile hazards, the staff noted that the licensee's strategy of relying 

on T-81 as a water source is an alternative to NE 12-06. Specifically NEI 12-06, Section 

3.2.1.3, states, in part, that water inventories contained in systems or structures with designs 

that are robust with respect to all applicable hazards are available. Although the licensee's 

strategy is an alternative to NEI 12-06, the staff finds it acceptable. See Section 3.14 for further 

discussion of the acceptability of this alternative. 

During the onsite audit, the licensee provided a calculation in EC-46465-08, "Freezing of 

Coolant Sources for FLEX Event." Rev. 4, which determined the time to reach freezing 

temperatures for tanks and stagnant hoses exposed to cold air during a BDBEE. The 

calculation showed that the CST would not freeze during the first 8 hours of the event and the T81 tanks will begin to freeze in 7.7 hours such that by the end of Phase 1 (8 hrs), the licensee 

showed that only 92 gal of the T-81 tank would be frozen. The FIP states that operators will 

begin staging the portable diesel driven FLEX pump 4 hours (2 hour duration) into the ELAP 

event. Given the conservative calculation assumptions the licensee made and the portable 

diesel driven FLEX pump will be staged before T-81 reaches a freezing temperature, the NRG 

staff finds it reasonable that both tanks should be available following an ELAP resulting from an 

extreme cold BDBEE. 

If implemented appropriately and consistent with the FIP, the licensee should have a water 

source available during the Phase 1 core cooling strategies for SG inventory makeup. In 

addition, the licensee's strategy should provide sufficient time for operators to deploy and stage 

Phase 2 FLEX equipment. The licensee's sequence of events timeline, as documented in FIP 

Table 1, shows that the CST and T-81 have sufficient water inventory for approximately 8 hours 

into the ELAP event. 

Phase 2 

The licensee states in its FIP, that the Phase 2 FLEX strategy for reactor core cooling and heat 

removal uses water from Lake Michigan for feeding the SGs using a portable diesel driven 

FLEX pump. The licensee plans to drop a suction hose through one of two removable grates 

into the Intake Canal. The staff noted that the supply of water from Lake Michigan to the intake 

canal is robust and is protected from all applicable hazards and is available to support Phase 2 

core cooling strategies. The licensee states in its FIP that in accordance with UFSAR Table 

5.2.2 and Section 5.5.1.1.2, the Intake Structure is robust with respect to all applicable external 

events and expected to be available following a BDBEE. 

Calculation EA-EC-46465-01, "Steam Generator Heat Transfer Capacity Using Lake Michigan," 

Rev. O, determined the effects on the heat transfer capabilities of the SGs when using Lake 

Michigan as a long-term makeup source. This analysis assumed that demineralized water is 

available for SG injection for the first eight hours after the event and then SG makeup will 
- 57 -

transition to Lake Michigan water with the required AFW flow rates. The results of this 

calculation show a 4.4 percent loss of heat transfer capability in the SGs 120 hours after the 

event. The NRC staff noted this minor loss in heat transfer capability is acceptable for operation 

up to 120 hours after the event due to the decrease in decay heat in the reactor that is required 

to be removed by the SGs. Approximately 72 hours after the event, a mobile water purification 

unit from the NSRC will be available. The analysis duration of 120 hours was chosen to provide 

margin beyond the 72 hours for deployment and operation of the water purification unit. The 

mobile purification unit will provide demineralized water to the FLEX pump for injection to the 

SGs such that the only potential for SG fouling (i.e., loss of heat transfer capability) would occur 

during the period when lake water is injected directly into the SGs without treatment. 

If implemented appropriately and consistent with the FIP, the licensee should have a water 

source available during the Phase 1 core cooling strategies for SG inventory makeup. In 

addition, the licensee's strategy should provide sufficient time for operators to deploy and stage 

Phase 2 FLEX equipment. The licensee's sequence of events timeline, as documented in FIP 

Table 1, shows that the CST and T-81 have sufficient water inventory for approximately 8 hours 

into the ELAP event. 

Phase 3 

The licensee's Phase 3 core cooling strategy does not rely on any additional water sources 

other than those discussed in Phase 2. However, the licensee does plan to use NSRC supplied 

portable water purifiers to provide cleaner water for SG makeup. 

3.10.2 Primary Coolant System Make-Up 

Phase 1 

In the FIP, Section 2.3.1 states that the RCS reactivity and inventory control is achieved initially 

by injection from the SITs as a result of RCS depressurization. In the Palisades UFSAR, Rev. 

29, Table 5.2-3 indicates that the SITs are Seismic Category I components. The FIP states that 

the SITs are located in the Reactor Building (Containment) which is described in UFSAR, Rev. 

21, Table 5.2-2 as a Class 1 structure; thus, protected from all applicable hazards. 

If implemented appropriately and consistent with the FIP, the licensee's approach should 

conserve RCS inventory to preclude the necessity for RCS system makeup during Phase 1. 

However, should makeup be necessary, the NRC staff finds that the SITs and associated piping 

and valves are robust and are expected to be available during ELAP event. 

Phase 2 and 3 

For PCS inventory control in Phase 2, PNP will utilize one of its two installed (FLEX portable 

diesel re-powered) positive displacement charging pumps, with borated water supplied from the 

BASTs. In its FIP, the licensee states the BASTs are estimated to provide PCS inventory 

control for greater than 24 hours after the event. Prior to depletion of the BASTs, boric acid 

batching operations will begin using the installed boric acid batching tank (T-77). Water for 

boric acid batching is supplied from the portable diesel driven FLEX pump. Hoses will be routed 

from the FLEX pump distribution manifold to tank T-77. During the audit process, the licensee 

provided report-RPT-13-00051, "Boric Acid Storage Tank (BAST) Suction Piping and 
- 58 -

Component Evaluation for FLEX," Rev. 0. The report found the BASTs, T-77, and all related 

piping to be robust for all applicable external hazards. If implemented appropriately and 

consistent with its FIP, the licensee should have a sufficient source of water available during 

Phase 2 to maintain RCS inventory in order to maintain natural circulation cooling and control 

reactivity in the core. 

3.10.3 Spent Fuel Pool Make-Up 

The licensee stated in its FIP that SFP cooling through makeup and spray will be provided by 

using the portable diesel driven FLEX pump with water from Lake Michigan, as described in SE 

Section 3.10.1. 

3.10.4 Containment Cooling 

The licensee stated in its FIP that no containment cooling is required for 10 days into the event 

at which time equipment from the NSRC will be available. 

As discussed in Section 3.4.3, the Containment analysis, EA-EC46465-02 showed that the 

CACs are not required for over 120 hours for the at-power scenario. During the plant recovery 

phase (i.e. Phase 3) the CACs may be utilized, although they are not credited for the FLEX 

coping strategy. During the recovery phase, the CACs may provide cooling water via the Phase 

3 NSRC UHS pump and the SWS tie-in location to provide cooling water for necessary plant 

heat loads including containment heat removal. 

3.10.5 Conclusions 

Based on the evaluation above, the NRC staff concludes that the licensee has developed 

guidance that, if implemented appropriately, should maintain satisfactory water sources 

following a BDBEE consistent with NEI 12-06 guidance, as endorsed by JLD-ISG-2012-01, and 

should adequately address the requirements of the order. 

3.11 Shutdown and Refueling Analyses 

Order EA-12-049 requires that licensees must be capable of implementing the mitigation 

strategies in all modes. In general, the discussion above focuses on an ELAP occurring during 

power operations. This is appropriate, as plants typically operate at power for 90 percent or 

more of the year. When the ELAP occurs with the plant at power, the mitigation strategy initially 

focuses on the use of the steam-driven TDAFW pump to provide the water initially needed for 

decay heat removal. If the plant has been shut down and all or most of the fuel has been 

removed from the RPV and placed in the SFP, there may be a shorter timeline to implement the 

makeup of water to the SFP. However, this is balanced by the fact that if immediate cooling is 

not required for the fuel in the reactor vessel, the operators can concentrate on providing 

makeup to the SFP. The licensee's analysis shows that following a full core offload to the SFP, 

about 3.28 hours are available to implement makeup before boil-off results in the water level in 

the SFP dropping far enough to uncover fuel assemblies, and the licensee has stated that they 

have the ability to implement makeup to the SFP within that time. 
- 59 -

When a plant is in a shutdown mode in which steam is not available to operate the steampowered pump and allow operators to release steam from the SGs (which typically occurs when 

the PCS has been cooled below about 300°F), another strategy must be used for decay heat 

removal. On September 18, 2013, NEI submitted to the NRC a position paper entitled 

"Shutdown/Refueling Modes" [Reference 25], which described methods to ensure plant safety in 

those shutdown modes. By letter dated September 30, 2013 [Reference 26], the NRC staff 

endorsed this position paper as a means of meeting the requirements of the order. 

The position paper provides guidance to licensees for reducing shutdown risk by incorporating 

FLEX equipment in the shutdown risk process and procedures. Considerations in the shutdown 

risk assessment process include maintaining necessary FLEX equipment readily available and 

potentially pre-deploying or pre-staging equipment to support maintaining or restoring key safety 

functions in the event of a loss of shutdown cooling. The NRC staff concludes that the position 

paper provides an acceptable approach for demonstrating that the licensees are capable of 

implementing mitigating strategies in shutdown and refueling modes of operation. In Section 

2.16 of the Fl P, the licensee described the procedural revisions that have been made to follow 

the guidance in the position paper. 

Containment Analyses 

Review of the scenario for Modes 5 and 6 indicates containment venting will be required to 

prevent exceeding containment design conditions. 

The NRC staff reviewed calculation EA-EC46465-02, "MAAP Containment Analysis for 

BDBEE," Rev. 0. This calculation used the MAAP computer code, version 4.0.6. The Mode 5 

and 6 analysis assumes that the plant has been shut down for 48 hours prior to the event, the 

PCS is at 200°F, 14.7 psia, and is at a reduced inventory (i.e., mid-loop). The PCS makeup is 

assumed available to prevent the core from being uncovered. As the PCS heats up and begins 

to boil, decay heat removal occurs by steam transfer into Containment. Without any mitigating 

actions in this scenario, Containment design limits of temperature and pressure were exceeded 

14.6 hours after the event. The analysis also determined that by opening a 3.75-in. equivalent 

diameter penetration in Containment, the Containment temperature and pressure design limits 

are not exceeded for the analysis duration of 120 hours. 

The analysis assumed that the action to open a Containment vent path would be taken after a 5 

psig rise in Containment pressure following the event. Either one of two 8-in Containment purge 

headers (JBB-1-8 in and JBB-2-8 in) can be used to provide Containment venting. Two AOVs 

in either line (CV-1805 and CV-1806 on JBB-2-8 in; CV-1807 and CV-1808 on JBB-1-8 in) must 

be held open to permit flow. The air then discharges into a filter housing on the 607' elevation 

of the CCW Room. A door on the CCW filter will be opened to allow the Containment 

atmosphere to exhaust to the CCW Room. A double door on the 625-ft elevation of the CCW 

Room is also opened allowing the room to exhaust directly to the atmosphere completing the 

vent path. Nitrogen is supplied from Nitrogen Backup Station 9 on the 625-ft Auxiliary Bay Roof 

to the AOVs. In addition, the portable electric driven FLEX air compressor provides flexibility 

and can be used to supply the actuators of the valves if staged and available. The purge valves 

are initially opened with the nitrogen supply and then mechanically gagged in the open position 

to facilitate the venting of Containment. 
- 60 -

Procedure ECN-56388 defines the containment vent path and guidance. Guideline FSG-15 

(Station Blackout While Shutdown) provides actions necessary to ensure a controlled process is 

in place for containment venting. 

The maximum Containment pressure reached was calculated to be 36.2 psig (at 58.98 hours) 

and the maximum Containment temperature reached was 281.4°F (at 58.98 hours). The 

Containment building remains below the internal design limits of 55 psig and 283°F. 

Based on the licensee's incorporation of the use of FLEX equipment in the shutdown risk 

process and procedures, the NRC staff concludes that the licensee has developed guidance 

that if implemented appropriately should maintain or restore core cooling, SFP cooling, and 

containment following a BDBEE in shutdown and refueling modes consistent with NEI 12-06 

guidance, as endorsed by JLD-ISG-2012-01, and should adequately address the requirements 

of the order. 

3.12 Procedures and Training 

The licensee stated in its FIP that the inability to predict actual plant conditions that require the 

use of FLEX equipment makes it impossible to provide specific procedural guidance. As such, 

the FSGs will provide guidance that can be employed for a variety of conditions. Clear criteria 

for entry into FSGs will ensure that FLEX strategies are used only as directed for BDBEE 

conditions, and are not used inappropriately in lieu of existing procedures. When FLEX 

equipment is needed to supplement EOP or abnormal operating procedure (AOP) strategies, 

the EOP or AOP, severe accident mitigation guidelines (SAMGs), or extreme damage mitigation 

guidelines (EDGMs) will direct the entry into and exit from the appropriate FSG procedure. 

FLEX support guidelines will provide available, pre-planned FLEX strategies for accomplishing 

specific tasks in the EOPs or AOPs. The FSGs will be used to supplement (not replace) the 

existing procedure structure that establishes command and control for the event. Procedural 

Interfaces have been incorporated into Procedure EOP-3.0, "Station Blackout Recovery" to the 

extent necessary to include appropriate reference to FSGs and provide command and control 

for the ELAP. Additionally, procedural interfaces have been incorporated into AOP-38 "Acts of 

Nature" to include appropriate reference to FSGs: 

The FSG maintenance will be performed by the site procedures group. In accordance with site 

administrative procedures, NEI 96-07, Rev. 1, "Guidelines for 10 CFR 50.59 Implementation," 

and NEI 97-04, Rev. 1, "Design Bases Program Guidelines," are to be used to evaluate 

changes to current procedures, including the FSGs, to determine the need for prior NRC 

approval. However, per the guidance and examples provided in NEI 96-07, Rev. 1, changes to 

procedures (EOPs, AOPs, EDMGs, SAMGs, or FSGs) that perform actions in response to 

events that exceed a site's design-basis should screen out. Therefore, procedure steps which 

recognize the ELAP/LUHS has occurred and which direct actions to ensure core cooling, SFP 

cooling, or Containment integrity should not require prior NRC approval. FSGs will be reviewed 

and validated by the involved groups to the extent necessary to ensure the strategy is feasible. 

Regarding training, the licensee stated in its FIP, that Entergy's nuclear training program has 

been revised to ensure that personnel proficiency in the mitigation of BDBEEs is adequate and 

maintained. These programs and controls were developed and have been implemented in 

accordance with the systematic approach to training (SAT) process. Initial training has been 
- 61 -

provided and periodic training will be provided to site emergency response leaders on BDBEE 

emergency response strategies and implementing guidelines. Personnel assigned to direct the 

execution of mitigation strategies for BDBEEs have received the necessary training to ensure 

familiarity with the associated tasks, considering available job aids, instructions, and mitigating 

strategy time constraints. Care has been taken to not give undue weight (in comparison with 

other training requirements) for operator training for BDBEE accident mitigation. The 

testing/evaluation of operator knowledge and skills in this area has been similarly weighted. Per 

American National Standards Institute/American Nuclear Society 3.5, nuclear power plant 

simulators for use in operator training, certification of simulator fidelity is considered to be 

sufficient for the initial stages of the BDBEE scenario until the current capability of the simulator 

model is exceeded. Full scope simulator models will not be upgraded to accommodate FLEX 

training or drills. 

The NRC staff finds that the licensee has adequately addressed the procedures and training 

associated with FLEX. The procedures have been issued and a training program has been 

established that will be maintained in accordance with NEI 12-06, Section 11.6. 

3.13 Maintenance and Testing of FLEX Equipment 

As a generic issue, NEI submitted a letter dated October 3, 2013, which included EPRI 

Technical Report 3002000623, "Nuclear Maintenance Applications Center: Preventive 

Maintenance Basis for FLEX Equipment." By letter dated October 7, 2013, the NRC endorsed 

the use of the EPRI report and the EPRI database as providing a useful input for licensees to 

use in developing their maintenance and testing programs. Preventative maintenance (PM) 

templates for the major FLEX equipment including the portable diesel pumps and generators 

have also been issued. 

In its FIP, the licensee noted that maintenance and testing of FLEX equipment is governed by 

the Entergy PM Program as described in procedure EN-DC-324 and that the Entergy PM 

Program is consistent with Institute of Nuclear Power (INPO) INPO AP-913, "Equipment 

Reliability Process," and utilizes the EPRI PM Basis Database as an input in development of 

fleet specific Entergy PM Basis Templates. In those cases where EPRI templates were not 

available for the specific component types, the licensee developed PM actions based on 

manufacturer provided information/recommendations. 

Due to Palisades's unique equipment storage plan, the portable FLEX equipment may be 

unavailable for 45 days provided that the site FLEX capability (N) is available. This is an 

alternate approach to the guidance of NEI 12-06. If portable equipment becomes unavailable 

such that the site FLEX capability (N) is not maintained, the licensee would initiate actions within 

24 hours to restore the site FLEX capability (N) and implement compensatory measures (e.g., 

repair equipment, use of alternate suitable equipment or supplemental personnel) within 72 

hours. 

The NRC staff finds that the licensee has adequately addressed equipment maintenance and 

testing activities associated with FLEX equipment because a maintenance and testing program 

has been established in accordance with NEI 12-06, Section 11.5. 
- 62 -

3.14 Alternatives to NEI 12-06. Revision O 

3.14.1 Use of Installed Charging Pumps 

The licensee states in its FIP that the strategy of repowering installed equipment is allowed per 

Table D-1 of NEI 12-06. However, the licensee also recognizes that reliance on only installed 

equipment during Phase 2 for RPV/PCS/SG makeup conflicts with the guidance provided in NEI 

12-06 Section 3.2.2(13). While the utilization of the installed charging pumps, with no portable 

equipment capability, does not meet the guidance of N El 12-06, it is considered an alternate 

method to comply with the order. The NRG staff noted that the licensee's strategy involves: 

• Two 100%-capacity, redundant pumps. 

• One pump providing 44 gpm, which exceeds the flow requirement of 30 gpm 

from the licensee's calculations. 

• Powering either pump from a separate LCC. 

• Diverse injection paths from either charging pump discharge piping are 

provided through either PCS loop injection piping or through HPSI piping. 

• A primary and alternate strategy for receiving power from the FLEX 

generator. The selected pump could be powered from the portable Phase 2 

FLEX generator through either of two LCCs which can be cross-connected so 

that either charging pump can be powered by either connection point. 

The NRG staff finds that the licensee has a strategy to provide PCS makeup that should prevent 

damage to the core following a BDBEE. Therefore, the NRG staff finds that the licensee's use 

of installed charging pumps is an acceptable alternative to NEI 12-06, and should meet the 

requirements of Order EA-12-049. 

3.14.2 FLEX Equipment Outage Times 

The Palisades FLEX storage configuration utilizes two storage locations; one located inside the 

PA and one located outside the PA. FLEX storage building "A" is a pre-engineered metal 

building located inside the PA. FLEX storage building "B" is a hardened reinforced concrete 

building located outside the PA. In NEI 12-06, Rev. 0, Section 10.1, stipulates that N+ 1 sets of 

FLEX equipment be protected from all applicable BDBEEs. While each FLEX storage building 

("A" and "B") stores a complete set of FLEX equipment, including debris removal equipment, 

both storage locations are required to ensure that a single set (N) of FLEX equipment would be 

available for all applicable hazards. Therefore, the Palaisades FLEX storage configuration does 

not protect N+ 1 sets of FLEX equipment from all applicable BDBEEs as stipulated in NEI 12-06, 

Rev. 0, Section 10.1. 

In its FIP, the licensee acknowledges that the Palisdaes FLEX storage configuration is an 

alternative to the guidance in NEI 12-06, Rev. 0. The FIP also describes a reduced allowed out 

of service time for FLEX equipment maintenance to address the alternative FLEX storage 

configuration. The NRG staff notes that by letter dated December 10, 2015, NEI submitted 

guidance document NEI 12-06, Rev. 2 [Reference 28] to the NRG for review. By letter dated 

January 22, 2016 [Reference 29], the NRG staff endorsed NEI 12-06, Rev. 2. Guidance 

document NEI 12-06, Rev. 2, contains modifications which resulted in NRG acceptance of the 

storage of backup (N+ 1) equipment such that it is not protected from the applicable BDBEE 
- 63 -

hazards. Section 11.5.4.b of NEI 12-06, Revision 2, contains the condition that if the site FLEX 

capability (N) is met, but not protected for all of the site's applicable hazards, then the allowed 

unavailability is reduced to 45 days (compared to the 90 day unavailability with any FLEX 

equipment unavailable, but with the FLEX capability (N) available and in a protected or diverse 

storage configuration). Although Palisades is evaluated to NEI 12-06, Rev. 0, in this SE, the 

licensee has committed to follow the 45 day unavailability limit stated in NEI 12-06, Rev. 2. 

Therefore, the NRG staff finds the Palisades storage configuration acceptable. 

The NRG staff reviewed the licensee's proposal and finds that the methods used to ensure that 

the primary (N) set of equipment is available, with a reduction in allowed unavailability to 45 

days if any N equipment is not protected for all of the site's applicable hazards, is an acceptable 

alternative to the NEI 12-06, and should meet the requirements of Order EA-12-049. 

3.14.3 Tank T-81 Tornado Missile Protection 

The licensee stated in its FIP that the method of protection for tornado generated missiles for 

tank T-81 is an alternate approach to the guidance of NEI 12-06. Specifically Section 2.1.1 of 

the FIP identifies that water inventories are available that are contained in systems or structures 

with designs that are robust with respect to all applicable hazards. Tank is fully protected from 

all design-basis missile hazards with the exception of the large missile hazard represented by 

an automobile. The licensee notes in Section 2.3.4.4 of its FIP that reasonable protection from 

the automobile missile hazard is provided by the following: 

• Surrounding structures that intervene and provide protection from the design 

basis automobile-like missiles; 

• Low likelihood of historic tornadic activity at the site/region; 

• Administrative controls in AOP-38, "Acts Of Nature," Rev. 4, to remove any 

potential automobile-like missiles from the zone of influence given the onset 

of high winds; 

• Implementation of modifications to restrain existing structures within the zone 

from becoming automobile like hazards; and 

• Administrative controls in FSGs that take action to provide alternative suction 

inventory from one of several non-robust sources within 3.5 hours of the 

onset of an ELAP if T-81 is determined to be compromised. 

During the audit, the licensee provided the evaluation, "Tornado Generated Missile Hazard 

Evaluation for the Primary System Makeup Storage Tank (T-81 ),"Rev. 2. This evaluation 

discussed in detail the above approach to providing automobile-like missile protection for T-81. 

While onsite, the NRG staff walked down the location of the T-81 tank and noted that the layout, 

intervening structures and potential automobile-like missiles were consistent with the licensee's 

evaluation. Given the procedural guidance the licensee has in place, the surrounding 

intervening structures and the additional protective measures, the NRG staff finds that the 

licensee's use of the above method to protect T-81 from an large, tornado-generated missile is 

an acceptable alternative to NEI 12-06, and that the licensee should meet the requirements of 

Order EA-12-049. 
- 64 -

3.15 Conclusions for Order EA-12-049 

Based on the evaluations above, the NRC staff concludes that the licensee has developed 

guidance to maintain or restore core cooling, SFP cooling, and containment following a BDBEE 

which, if implemented appropriately, should adequately address the requirements of Order EA12-049. 

4.0 TECHNICAL EVALUATION OF ORDER EA-12-051 

By letter dated February 28, 2013 [Reference 30], the licensee submitted its OIP for PNP in 

response to Order EA-12-051. By letter dated July 18, 2013 [Reference 31] the NRC staff sent 

a RAI to the licensee. The licensee provided a response by letter dated August 19, 2013 

[Reference 32]. By letter dated November 26, 2013 [Reference 33], the NRC staff issued an 

ISE and RAI to the licensee. 

By letters dated August 28, 2013 [Reference 34], February 28, 2014 [Reference 35], August 28, 

2014 [Reference 36], February 27, 2015 [Reference 37], and August 28, 2015 [Reference 38], 

the licensee submitted status reports for the Integrated Plan. The Integrated Plan describes the 

strategies and guidance to be implemented by the licensee for the installation of reliable SFP 

level instrumentation, which will function following a BDBEE, including modifications necessary 

to support this implementation, pursuant to Order EA-12-051. By letter dated December 16, 

2015 [Reference 39], the licensee reported that full compliance with the requirements of Order 

EA-12-051 was achieved. 

The licensee has installed a SFP level instrumentation system designed by Mohr Test and 

Measurement LLC. The NRC staff reviewed the vendor's SFP level instrumentation system 

design specifications, calculations and analyses, test plans, and test reports. The staff issued 

an audit report on August 27, 2014 [Reference 40]. 

The staff performed an onsite audit to review the implementation of SFP level instrumentation 

related to Order EA-12-051. The scope of the audit includes verification of (a) site's seismic and 

environmental conditions enveloped by the equipment qualifications, (b) equipment installation 

met the requirements and vendor's recommendations, and (c) program features met the 

requirements. By letter dated October 13, 2015 [Reference 18], the NRC issued an audit report 

on the licensee's progress. Refer to Section 2.2 above for the regulatory background for this 

section. 

4.1 Levels of Required Monitoring 

In its OIP, the licensee stated that Level 1 is elevation 645 ft. 7 in. Level 1 is the level adequate 

to support operation of the normal fuel pool cooling system. It is the higher of the level at which 

reliable suction loss occurs due to uncovering the coolant inlet pipe or any weirs or vacuum 

breakers associated with suction loss. Level 2 is elevation 634 ft. 5 in. (1 Oft. above Level 3). 

Level 2 is the level adequate to provide substantial radiation shielding for a person standing on 

the spent fuel pool operating deck. Level 3 is elevation 624 ft. 5 in. Level 3 is the level where 

fuel remains covered. It is defined as the highest point of any fuel rack seated in the spent fuel 

pool (within ± 1 ft.). 
- 65 -

In its letter dated August 19, 2013 [Reference 32], the licensee provided a sketch depicting that 

Level 1 is designated at 645 ft. - 7 in. Elevation, Level 2 is designated at 634 ft. - 5 in. Elevation, 

and Level 3 is designated at 625 ft. - 5 in. Elevation. The sketch also shows the measurement 

range for the spent fuel pool instrumentation (SFPI) is between 625 ft. - 2 in. and 647 ft. - 1 O in. 

elevations. 

The NRC staff noted that the designated Level 1 (elevation 645 ft. - 7 in.) is adequate for normal 

SFP cooling system operation and it is also adequate to ensure the required fuel pool cooling 

pump net positive suction head. This level also represents the higher of the two points 

described in NEI 12-02, Section 2.3.1 for Level 1. The designated Level 2 (elevation 634 ft. 5 

in.) uses the first of the two options described in NEI 12-02 tor Level 2, which is 1 Oft. ± 1 ft. 

above the highest point of any fuel rack seated in the spent fuel pool. The designated Level 3 

(elevation 625 ft. - 5 in.) is 1 ft. above the highest point of any spent fuel storage rack seated in 

the SFP. 

The NRC staff finds that the licensee's proposed Levels 1, 2 and 3 are consistent with NEI 12-

02 guidance, as endorsed by JLD-ISG-2012-03, and adequately addresses the requirements of 

the order. 

4.2 Evaluation of Design Features 

Order EA-12-051 required that the SFP level instrumentation shall include specific design 

features, including specifications on the instruments, arrangement, mounting, qualification, 

independence, power supplies, accuracy, testing, and display. Refer to Section 2.2 above for 

the requirements of the order in regards to the design features. Below is the NRC staff's 

assessment of the design features of the SFP level instrumentation. 

4.2.1 Design Features: Instruments 

In its OIP, the licensee stated that the primary instrument channel (Channel A) and backup 

instrument channel (Channel B) are permanent, fixed channels. The instrument channels will 

provide level indication through the use of Guided Wave Radar (GWR) technology through the 

principle of Time Domain Reflectometry (TOR). The instrument provides a single continuous 

span from above Level 1 to within 1 ft. of the top of the spent fuel racks. 

In its OIP, the licensee also provided a sketch depicting the elevations identified as Levels 1, 2 

and 3 and the SFP level instrumentation measurement range. The NRC staff noted that the 

measurement range goes from the 647 ft. - 10 in. to 625 ft. - 2 in. This measurement range will 

cover Levels 1, 2, and 3 as described in Section 4.1 above. 

The NRC staff finds that the licensee's design, with respect to the number of channels and 

measurement range for its SFP, appears to be consistent with NEI 12-02 guidance, as 

endorsed by JLD-ISG-2012-03, and should adequately address the requirements of the order. 

4.2.2 Design Features: Arrangement 

In its OIP, the licensee stated that the location of the Channel A display/processor is in the 

MCR, and the location of the Channel B display/processor is in the Reactor Auxiliary Building. 

These locations are expected to be a mild environment after a BDBEE and can be easily 
- 66 -

accessed from the MGR; therefore, personnel can promptly obtain readings from the display. 

These areas provide adequate protection against the effects of temperature, flood, humidity, 

radiation, seismic events, and missile hazards. 

In its OIP, the licensee also provided a sketch depicting the SFP instrument locations and 

conduit routings of both primary and back-up level instrument channels. The NRG staff noted 

that, and verified by walkdown during the onsite audit, that there is sufficient channel separation 

within the SFP area between the primary and back-up level instruments, sensor electronics, and 

routing cables to provide reasonable protection of the level indication function against missiles 

that may result from damage to the structure over the SFP. 

The NRG staff finds that the licensee's arrangement for the SFP level instrumentation, if 

implemented appropriately, appears to be consistent with NEI 12-02 guidance, as endorsed by 

JLD-ISG-2012-03, and should adequately address the requirements of the order. 

4.2.3 Design Features: Mounting 

In its letter dated August 19, 2013 [Reference 32], the licensee stated that the loading on the 

probe mount and probe body includes both seismic and hydrodynamic loading using seismic 

response spectra that bound the PNP design basis maximum seismic loads applicable to the 

installation location(s). Seismic loading response of the probe and mount is separately modeled 

using finite element modeling software (GOTHIC). The GOTHIC-derived fluid motion profile in 

the pool at the installation site and resultant distributed hydrodynamic loading terms are added 

to the calculated seismic loading terms in the finite element model to provide a conservative 

estimate of the combined seismic and hydrodynamic loading terms for the probe and probe 

mount, specific to the chosen installation location for the probe. The proximal portion of the 

level probe is designed to be attached near its upper end to a Seismic Category I mounting 

bracket configured to suit the requirements of a particular SFP. The bracket may be bolted 

and/or welded to the SFP deck and/or SFP liner/wall according to the requirements of the 

particular installation per Seismic Category I requirements. 

In its letter dated February 27, 2015 [Reference 37], the licensee further stated that the site 

specific seismic analysis bounds PNP's seismic criteria. The qualification report envelops all 

components of the new SFP level instrumentation required to be operational during a BDBEE 

and post-event. Calculations EA-EC46466-01 and EA-EC46466-02 determine that all 

components, supports, and anchorages required are structurally adequate and seismically 

qualified as all Interaction Ratios (IR) are less than one (1.0). The seismic qualification reports 

in combination with NAl-1725-003, NAl-1725-004, and PNP site specific Report NAl-1791-006 

adequately bound the hydrodynamic loads associated with sloshing for PNP. Calculation EAEC46466-01 also accounts for sloshing effects to the probe. 

The NRG staff noted that the licensee adequately addressed the design criteria and 

methodology used to estimate and test the total loading on the mounting devices, including the 

design basis maximum seismic loads and the hydrodynamic loads that could result from pool 

sloshing. The site-specific seismic analyses demonstrated that the SFP level instrumentation's 

mounting design is satisfactory to allow the instrument to function per design following the 

maximum seismic ground motion. The NRG staff verified the instrument channel's mounting 

design by performing a walkdown and by reviewing the following: 
- 67 -

• Calculation EA-EC46466-01, "Design of SFPI Probe Mounting Bracket," Rev. O 

• Calculation EA-EC46466-02, "Qualification of SFP Instrumentation Mounting Details," 

Rev. 0 

• NAl-1791-006, "Seismic Induced Hydraulic Response in the Palisades Spent Fuel Pool," 

Rev. 0 

Based on the discussion above, the NRG staff finds the licensee's proposed mounting design 

appears to be consistent with NEI 12-02 guidance, as endorsed by JLD-ISG-2012-03, and 

should adequately address the requirements of the order. 

4.2.4 Design Features: Qualification 

4.2.4.1 Augmented Quality Process 

Appendix A-1 of the guidance in NEI 12-02 describes a quality assurance process tor nonsafety systems and equipment that is not already covered by existing quality assurance 

requirements. Per JLD-ISG-2012-03, the NRG staff found the use of this quality assurance 

process to be an acceptable means of meeting the augmented quality requirements of Order 

EA-12-051. 

In its OIP, the licensee stated that augmented quality requirements will be applied to all 

components in the instrumentation channels for: 

• design control 

• procurement document control 

• instructions, procedures, and drawings 

• control of purchased material, equipment, and services 

• inspection, testing, and test control 

• inspections, test, and operating status 

• nonconforming items 

• corrective actions 

• records 

• audits 

In its letter dated December 16, 2015 [Reference 39], the licensee further stated that the SFP 

instrumentation channels have been designated as augmented quality per Entergy processes 

covering procurement, design, and installation. Per Entergy procedures, augmented quality 

refers to SSCs that have been designated as requiring additional quality level oversight or 

requirements beyond what would be required for non-safety related SSCs. As such, the SFP 

instrument channels have demonstrated reliability through establishment of an augmented 

quality processes at applicable environmental extremes. 

The NRG staff noted that the licensee's proposed augmented quality assurance process 

appears to be consistent with NEI 12-02, as endorsed by JLD-ISG-2012-03. If implemented 

appropriately, this approach appears to be consistent with NEI 12-02 guidance, and should 

adequately address the requirements of the order. 
- 68 -

4.2.4.2 Equipment Reliability 

The NRC staff reviewed the Mohr SFP level instrumentation's qualification and testing during 

the vendor audit for temperature, humidity, radiation, shock and vibration, and seismic 

[Reference 40]. The staff further reviewed the anticipated Palisades' environmental conditions 

during the on-site audit [Reference 18]. 

4.2.4.2.1 Temperature. humidity, and radiation 

In its OIP, the licensee stated that components in the area of the SFP will be designed for the 

temperature, humidity, and radiation levels expected during normal, event, and post-event 

conditions for no fewer than seven days post-event or until off-site resources can be deployed 

by the mitigating strategies resulting from Order EA-12-049. Equipment located in the SFP will 

be qualified to withstand a total accumulated dose of expected lifetime at normal conditions plus 

accident dose received at post event conditions with SFP water level within 1 foot of the top of 

the fuel rack seated in the spent fuel pool. The probe and cable in the spent fuel pool area are 

robust components that are not adversely affected by expected radiation, temperature, or 

humidity. The areas selected for display/processor installation are considered mild 

environments. 

In its letter dated August 19, 2013 [Reference 32], the licensee stated that the signal processor 

is designed for mild environment with radiation levels similar to background radiation. Physical 

testing in an environmental chamber to demonstrate normal operation at the operating 

temperatures and humidity specified for the instrument. The probe assembly is qualified by 

materials properties and use history of substantially similar probe designs in SG applications at 

significantly higher temperatures and pressures and saturated steam environments. 

In its letter dated February 27, 2015 [Reference 37], the licensee further stated that the 

Palisades Control Room (Channel A display's location) and the C-40 Panel Room (Channel B 

display's location) are considered a mild environment. Radiation levels in the Control Room and 

C-40 Panel Room are not impacted by a reduction in SFP water level. Calculation EAEC46465-03 determined that the maximum temperature in the control room will be 121°F. The 

operating temperature of the C-40 Panel Room is 50-110°F. Humidity in the Control Room and 

C-40 Panel Room are normally regulated by the HVAC system at 50 percent. The SFPI system 

electronics was tested to a nominal temperature range of 14°F to 131°F and a humidity range of 

5 percent to 95 percent relative humidity (RH). During an extended loss of ac power, the HVAC 

system is no longer available. Assuming the Control Room and C-40 Panel Room remain 

isolated from outside air, the temperature is expected to increase and the humidity is expected 

to decrease because the heat loads are dominated by the sensible heat of electrical equipment. 

Therefore, the maximum temperatures of 121°F for the Control Room and 110°F for the C-40 

Panel Room at a humidity below 50 percent is bounded by results from the test case of 55°C 

(131°F), which is the 50 percent RH. 

The NRC staff found that the licensee adequately addressed the equipment reliability of SFP 

level instrumentation with respect to temperature, humidity and radiation. The equipment 

qualification envelops the expected Palisades' radiation, temperature, and humidity conditions 

during a BDBEE. The equipment environmental testing demonstrated that the SFP 

instrumentation should maintain its functionality during the expected BDBEE conditions. 
- 69 -

4.2.4.2.2 Shock and Vibration 

In its letter dated February 27, 2015 [Reference 37], the licensee stated that the probe and 

repairable head are considered inherently resistant to shock and vibration. They are evaluated 

to be adequately designed for resilience against shock and vibration. The probe mounting 

components and fasteners are designed as rigid components inherently resistant to vibration 

effects. The probes will be affixed to the bracket using a machine screw connection designed 

with proper thread engagement and lock washers. The indicator and battery enclosures will be 

mounted in the Control Room and C-40 Panel Room. The equipment is not affixed or adjacent 

to any rotating machinery that would cause vibration effects in the area of installation. The 

instrument mounting components and fasteners are designed as rigid components inherently 

resistant to vibration effects. 

The NRC staff found that the licensee adequately addressed the equipment reliability of SFP 

level instrumentation with respect to shock and vibration qualifications. The staff also reviewed 

the shock and vibration test report during the vendor audit and found it acceptable. 

4.2.4.2.3 Seismic 

In its letter dated August 19, 2013 [Reference 32], the licensee stated that triaxle shake-table 

testing is planned to be performed for the signal processor to envelope Seismic Category I SSE 

conditions or Palisades design-basis maximum seismic loads using Institute of Electrical and 

Electronics Engineers (IEEE) 344-2004 methodology. A seismic and hydrodynamic finite 

element analysis is performed by the vendor using enveloping Seismic Category I SSE 

conditions or Palisades design-basis maximum seismic loads relative to the location where the 

equipment is mounted. Combined seismic and hydrodynamic analysis will be used to 

demonstrate that the probe waveguide's geometric dimensions do not change significantly as a 

result of the seismic conditions. 

In its letter dated February 27, 2015 [Reference 37], the licensee further stated that the site 

specific seismic analysis bounds Palisades seismic criteria. The qualification report envelops all 

components of the new SFP level required to be operational during BDBEE post-event. 

Calculations EA-EC46466-01 and EA-EC46466-02 determine that all components, supports, 

and anchorages required are structurally adequate and seismically qualified as all IRs are less 

than one (1.0). 

The NRC staff found that the licensee adequately addressed the equipment reliability of SFP 

level instrumentation with respect to seismic qualification. The equipment's seismic 

qualifications envelop the expected Palisades' seismic condition during a BDBEE. The sitespecific seismic analyses demonstrated that the SFP level instrumentation's mounting design is 

satisfactory to allow the instrument to function per design following the maximum seismic 

ground motion. Further discussion of the instrument channel's mounting design is described in 

Subsection 4.2.3, "Design Features: Mounting". The NRC staff reviewed the vendor's factory 

acceptance test reports and found the Palisades SFPI design and qualification process 

acceptable. However, during the onsite audit, the NRC staff learned that there were incidents at 

other nuclear facilities, in which the Mohr's SFPI equipment experienced failures of the filter coil 

(or choke). The staff requested Palisades to address the impact of SFPI equipment failures on 

the Palisades' SFP level instrument. 
- 70 -

In its letter dated August 28, 2015 [Reference 38], the licensee provided a response, in which it 

stated that the vendor determined that the source of the failures is a miniature surface mount 

common-mode choke component used on the Video and Digicomp printed circuit boards within 

the EFP-IL Signal Processor. Per the vendor's recommendation, the two boards have been 

replaced in both EFP-IL Signal Processors at Palisades. Corrective actions were performed 

under condition report CR-HON-2015-0345 to replace each of the two circuit boards within the 

two EFP-IL Signal Processors at the site. The new boards have equivalent substitute 

components that are less susceptible to transient electrical events. The substitute components 

have equivalent size, mass, and solder attachment technique as the original component such 

that there is no impact to the system mechanical characteristics. The components demonstrate 

equivalent electrical performance such that electromagnetic compatibility (EMC) characteristics 

are not significantly changed. 

During the onsite audit, the NRC staff reviewed Condition Report CR-HQN-2015-0345 and the 

Mohr Root Cause Analysis Report, Rev. 1 dated June 1, 2015. The NRC staff found Palisades' 

approach to address the equipment failure appropriate. However, the Mohr Root Cause 

Analysis Report did not adequately address the impact to order compliance. Section 6.2, 

"Impact to Regulatory Compliance," of the report stated, in part, that no impact to regulatory 

compliance was identified. On-going repairs are being carried out using more rugged but 

otherwise equivalent choke components with equivalent form, fit, and function using equivalent 

solder technique. Alteration of enclosure mass due to the repair is negligible (0.002 percent) 

and well within the expected variation of mass from enclosure to enclosure due to normal 

manufacturing variation of PCB assemblies, cables, and other components. Credit is therefore 

taken for previously documented regulatory qualification, including seismic qualification, per 

Order EA-120-051, NEI 12-02, Rev. 1 and JLD-ISG-2012-03. Repairs are being carried out in 

compliance with the quality requirements of NEI 12-02, Rev. 1, Appendix A-1, "Quality 

Assurance". 

The NRC staff found that the justifications for the repaired equipment's qualification with respect 

to seismic and EMC in the Mohr Root Cause Analysis Report are not sufficient. The report also 

did not address the shock, vibration, and environmental qualifications (temperature and 

humidity) for the replaced components. During the onsite audit, Palisades coordinated a 

conference call with Mohr Test and Measurement LLC. During the call, the NRC staff requested 

that Mohr provide a detailed description of what changes Mohr made to the SFPI equipment 

that have been sent to the Mohr facility for repair. The NRC staff also requested that Mohr 

provide the following qualifications for each replaced component/part to meet the Order EA-12-

051 requirements: 

• Seismic 

• Shock 

• Vibration 

• Temperature 

• Humidity 

• EMC 

Mohr issued Qualification Report: EFP-IL MOD 1 Modification Package, Rev. 0, dated July 16, 

2015, to document regulatory qualification of its SFPI equipment. In this report, Mohr provided 

evaluation of the following hardware modifications: 
- 71 -

• In-place replacement of the miniature T1 surface mount choke on the 01-

EFP-IL-50001 board with an equivalent component. 

• In-place replacement of the miniature T1 surface mount choke on the 01-

EFP-IL-50006 board with an equivalent component. 

• Incorporation of a fusible link in the 01-EFP-IL-50204 cable assembly. 

• Full electrical isolation added to the 01-EFP-IL-50007 (USB interface) board. 

Below is the summary of the evaluation of the above modifications: 

T1 Choke Replacement Evaluation 

Temperature and Humidity 

The replacement choke has an operating temperature range of -40 °C to +85 °C, 

exceeding the -10 °C to +55 °C requirement. Non-condensing humidity does not 

alter performance of this component. 

Electromagnetic Compatibility (EMC) 

There is no change to EMC qualification. The choke demonstrates equivalent or 

higher impedance to common mode noise. 

Shock and Vibration 

The mass differences are 0.002% and 0.47% for 01-EFP-IL-50001 enclosure 

mass and the board mass respectively and 0.0003% and 0.18% for 01-EFP-IL50006 enclosure mass and the board mass respectively. 

Seismic 

Qualification by similarity to existing qualified equipment is permitted by IEEE 

Std. 344-2004. Replacement of T1 choke does not significantly alter equipment 

mass, mass distribution, or other mechanical characteristics. 

Fusible Link Evaluation: 

The 01-EFP-IL-50204 Fusible Link is added to the existing power board power cable. One 

Fusible Link is used per EFP-IL signal processor. 

Temperature and Humidity 

The Fusible Link's fuses are rated for -55 °C to+ 125 °C and 100% RH per MILSTD-201 Method 106. 

Electromagnetic Compatibility (EMC) 

There is no change to EMC qualification. The Fusible Link uses insulated wiring 

and connectors identical in configuration to the remainder of the previously 

qualified cable assembly and expected emissions are unchanged. 
- 72 -

Shock and Vibration 

The Fusible Link uses insulated wiring and connectors identical in configuration 

to the remainder of the cable assembly and is secured using identical tie-down 

and strain-relief methods which have been previously qualified. 

The Fusible Link contributes approximately 0.14% enclosure mass, well within 

the expected variation of the unmodified EFP-IL signal processor enclosure mass 

due to manufacturing tolerances of system components. 

The connectors are qualified by the manufacturer for vibration conditions per EIA 

364-28 and shock loading at 50g. The connector rated minimum pull force is 8.0 

lbf per wire terminal, for a total rating of 80.0 lbf for the 1 O wire connector, 

equivalent to 2086g loading assuming Fusible Link mass of 17.4 g. 

The Littelfuse fuse lead axial pull force is rated at 7 lbs. per MIL-STD-202, which 

is equivalent to 182 g static loading per fuse (two fuses per cable), assuming 

cable mass of 17.4 g and conservatively neglecting stress shielding by cable 

wiring and insulation. 

Seismic 

The Fusible Link contributes approximately 0.14% enclosure mass, well within 

the expected variation of the unmodified EFP-IL signal processor enclosure mass 

due to manufacturing tolerances of system components. 

Electrical Isolation Evaluation: 

The front panel USB board 01-EFP-IL-50007 has been modified through addition of the 

component to provide galvanic USB isolation and enhance ESD protection (±15kV) 

Temperature and Humidity 

The additional component is rated for normal operation at -40 °C to +85 °C. The 

component is a hermetic plastic BGA package that is not susceptible to elevated 

humidity. 

Electromagnetic Compatibility (EMC) 

There is no change to EMC qualification. The isolation technology within the 

additional component is compliant with applicable standards including radiated 

emissions limit per IEC 61000/CISPR 22. The device reduces the equipment's 

already low radiated emissions when the USB device is in use because it isolates 

and prevents noise on internal data and power lines from propagating to external 

devices connected to the front-panel USP port. The device is not active when 

USB devices are not in use. 

Shock and Vibration 

There is insufficient mass difference to alter the equipment shock and vibration 

response characteristics. The additional component is a rugged, compact, 

encapsulated surface-mount BGA package enveloped in size and mass by other 

components in the system. Surface mount components as a class are not 

susceptible to required levels of shock and vibration when mounted within the 

EFP-IL equipment enclosures. 
- 73 -

Seismic 

The nominal difference in enclosure mass is trivial at 0.014, well within the 

expected variation of EFP-IL signal processor enclosure mass due to 

manufacturing tolerances of system components. 

The NRC staff found that the vendor adequately addressed the staff concern with regard to the 

modified equipment qualifications. The temperature and humidity ratings of the replacement 

parts envelop the expected Control Room's and C-40 Panel Room's conditions during a 

BDBEE. There is no indication that new electromagnetic emissions will be introduced by the 

replacement parts. The mass differences are insufficient to alter the seismic, shock, and 

vibration response characteristics. 

Based on the discussion above, the NRC staff finds that the licensee's proposed instrument 

qualification process appears to be consistent with NEI 12-02 guidance, as endorsed by JLDISG-2012-03, and should adequately address the requirements of the order. 

4.2.5 Design Features: Independence 

In its OIP, the licensee stated that the primary instrument channel will be independent of the 

backup instrument channel. Independence is obtained by physical separation of components 

between channels and the use of normal power supplied from separate 480V buses. 

Independence of power sources is described in Section 11. The 2 level instruments and any 

associated cabling (for each Unit SFP) will be physically separated and electrically independent 

of one another. 

In its letter dated August 19, 2013 [Reference 32], the licensee further stated that power for 

each channel is provided from independent 120Vac, 60 Hertz (Hz) power sources. Backup 

power is provided by a battery capable of providing continuous display operation for at least 3 

days. The battery will be provided with the display/processor. The design prevents failure of a 

single channel from causing the alternate channel to fail. Physical separation of the two 

channels will be accomplished by separately routing cable and conduit as much as practical. 

In its letter dated August 28, 2014 [Reference 36], the licensee further stated that instrument 

Channel A is being powered from Lighting Panel (L-25B) through existing 20A breaker #5, 

which is supplied from MCC #1. Instrument Channel B is being powered from Lighting Panel (L76) through existing 20A breaker #12, which is supplied from MCC #82. These two buses 

represent independent 480 V power sources. 

The NRC staff noted that the licensee adequately addressed the instrument channel 

independence. The primary instrument channel is physically and electrically independent of the 

backup instrument channel. The instrument channels' physical separation is discussed in 

Subsection 4.2.2, "Design Features: Arrangement". With the licensee's proposed power 

arrangement, the electrical functional performance of each level measurement channel would 

be considered independent of the other channel, and the loss of one power supply would not 

affect the operation of other independent channel under BDBEE conditions. 
- 74 -

Based on the discussion above, the NRC staff finds the licensee's proposed design, with 

respect to instrument channel independence, appears to be consistent with NEI 12-02 

guidance, as endorsed by JLD-ISG-2012-03, and should adequately address the requirements 

of the order. 

4.2.6 Design Features: Power Supplies 

In its OIP, the licensee stated that each instrument channel is normally powered from 120Vac 

60 Hz plant power to support continuous monitoring of SFP level. The primary channel receives 

power from a different 480V bus than the backup channel. Therefore, loss of any one 480V bus 

does not result in loss of normal 120 Vac power for both instrument channels. On loss of 

normal 120 Vac power, each channel's UPS automatically transfers to a dedicated backup 

battery. If normal power is restored, the channel will automatically transfers back to the normal 

ac power. The backup batteries are maintained in a charged state by commercial-grade 

uninterruptible power supplies. The batteries are sized to be capable of supporting intermittent 

monitoring for a minimum of 3 days of operation. This provides adequate time to allow the 

batteries to be replaced or until off-site resources can be deployed by the mitigating strategies 

resulting from Order EA-12-049. An external connection permits powering the system from any 

portable DC source. 

In its letter dated August 19, 2013 [Reference 32], the licensee further stated that permanent 

installed battery capacity for 7 days continuous operation is planned consistent with NEI 12-02 

duration without reliance on or crediting of potentially more rapid FLEX Program power 

restoration. Batteries are readily replaceable via spare stock without the need for recalibration 

to maintain accuracy of the instrument. These measures should ensure adequate power 

capacity and margin. 

During the onsite audit, the NRC staff inquired about the power restoration strategy following an 

ELAP and prior to the back-up battery depletion. The licensee provided a response, in which it 

stated that ac power is not credited for restoring power to the instrumentation in the event of an 

ELAP. Instead, external batteries will be used to supply power through the external connection 

ports along the bottom of the Battery Enclosure. Battery cables are stored in the storage area 

of the Control Room. This is captured in FSG-11, "Alternate SFP Makeup and Cooling," Rev. B. 

At least one battery per channel will kept on site and additional batteries can be taken from cars 

as needed. 

The NRC staff found the response acceptable. However, while verifying the response by 

reviewing FSG-11, the NRC staff noted that the battery's voltage was specified with different 

values (24 Vdc, 9-36 Vdc, and 12 Vdc). In response to the NRC staff's concern, the licensee 

issued Corrective Action CA-5 of CR-PLP-2015-02482 to clarify the verbiage within FSG-11 

concerning the proper and appropriate voltage source to provide an external source of power to 

the SFPI when normal and back-up power may be unavailable. 

Based on the discussion above, the NRC staff finds the licensee's proposed power supply 

design appears to be consistent with NEI 12-02 guidance, as endorsed by JLD-ISG-2012-03, 

and should adequately address the requirements of the order. 
- 75 -

4.2. 7 Design Features: Accuracy 

In its OIP, the licensee stated that the absolute system accuracy is better than± 3 in. This 

accuracy is applicable for normal conditions and the temperature, humidity, chemistry, and 

radiation levels expected for BDBEE conditions. 

In its letter dated August 19, 2013 [Reference 32], the licensee further stated that in general, 

relative to normal operating conditions, any applicable calibration procedure tolerances (or 

acceptance criterion) are planned to be established based on the manufacturer's 

stated/recommended reference accuracy (or design accuracy). The methodology used is 

planned to be captured in plant procedures and/or programs. 

The NRC staff found that instrument channel design accuracies and methodology sufficient for 

maintaining the instrument channels to within their designed accuracies before significant drift 

can occur. However, the NRC staff needed to verify that the channels will retain these accuracy 

performance values following a loss of power and subsequent restoration of power. 

In its letter dated February 27, 2015 [Reference 37], the licensee further stated that Report 1-

0410-10 concludes that the accuracy is not affected by an interruption in power. Report 1-0410-

8 concludes that the presence of borated water and/or boric acid deposits will not significantly 

impair the ability of the Mohr EFP-IL SFPI system to accurately measure water level in the SFP 

environment. Visual inspection and/or wash down of the probe assembly could be initiated by 

accuracy requirements or routine inspection. The probe head assembly includes a connection 

mechanism for flushing water to remove boron build-up as may be necessary. Alternatively, the 

SFP water level can be raised until it covers and dissolves the boric acid deposit. Appendix A of 

Report 1-0410-12 states that the absolute accuracy is 76.2 mm or 3.0 in, not including boric acid 

deposition effects. This error complies with the limit of ±1 foot set by NEI 12-02. Additionally, 

the probe is designed to produce accurate level indication in boiling and frothing (multiphase) 

environments according to Report 1-0410-15. 

The licensee has demonstrated that the instrument channels' accuracy is not significantly 

affected by BDBEE conditions. If implemented properly, the instrument channels should 

maintain the designed accuracy following a power source change or interruption without the 

need of recalibration. The NRC staff finds the licensee's proposed instrument accuracy appears 

to be consistent with NEI 12-02 guidance, as endorsed by JLD-ISG-2012-03, and should 

adequately address the requirements of the order. 

4.2.8 Design Features: Testing 

In its OIP, the licensee stated that the display/processor performs automatic in-situ calibration 

and automatically monitors for cable, connector, and probe faults using time domain 

reflectometry (TOR) technology. Channel degradation due to age or corrosion is not expected 

but can be identified by monitoring trends. 

In its letter dated August 19, 2013 [Reference 32], the licensee further stated that the level 

instrument automatically monitors the integrity of its level measurement system using in-situ 

capability. Periodic calibration checks of the signal processor electronics to extrinsic National 

Institute of Standards and Technology (NIST)-traceable standards can be achieved through the 

use of standard measurement and test equipment. It is planned to be periodically inspected 
- 76 -

electromagnetically using TOR at the probe hardline cable connector to demonstrate that the 

probe assembly meets manufactured specification and visually to demonstrate that there has 

been no mechanical deformation or fouling. 

Performance tests (functional checks) are automated and/or semi-automated (requiring limited 

operator interaction) and are performed through the instrument menu software and initiated by 

the operator. Other tests such as menu button tests, level alarm, and alarm relay tests are only 

initiated manually by the operator. Performance tests are planned to be performed periodically 

as recommended by the equipment vendor, for instance quarterly but no less often than the 

calibration interval of two years. 

Channel functional tests per operations procedures with limits established in consideration of 

vendor equipment specifications are planned to be performed at appropriate frequencies 

established equivalent to or more frequently than existing spent fuel pool instrumentation. 

Manual calibration and operator performance checks are planned to be performed in a periodic 

scheduled fashion with additional maintenance on an as-needed basis when flagged by the 

system's automated diagnostic testing features. Periodic (e.g., quarterly or monthly) review of 

the system level history and log files and routine attention to any warning message on the 

system display is recommended by the vendor. Formal calibration checks are recommended by 

the vendor on a 2-year interval to demonstrate calibration to external NIST-traceable standards. 

Formal calibration check surveillance interval and timing would be established consistent with 

applicable guidance (i.e., NEI 12-02, Section 4.3; on a refueling outage interval basis and within 

60 days of a planned refueling outage). Items such as system batteries are planned to be 

assessed under the preventive maintenance (PM) program for establishment of replacement 

frequency. Surveillance/PM timing/performance are planned to be controlled via tasks in the 

PM program. 

The NRC staff noted that the SFP level instrumentation is adequately designed to provide the 

capability for routine testing and calibration including in-situ testing/calibration. By comparing 

the levels in the instrument channels and the maximum level allowed deviation for the 

instrument channel design accuracy, the operators could determine if recalibration or 

troubleshooting is needed. The staff finds the licensee's proposed SFP instrumentation design 

allows for testing that appears to be consistent with NEI 12-02 guidance, as endorsed by JLDISG-2012-03, and should adequately address the requirements of the order. 

4.2.9 Design Features: Display 

In its OIP, the licensee stated that the primary and backup instrument displays will be located in 

the MCR and the Auxiliary Building. For both normal and expected BDBEE conditions, the 

displays are promptly accessible to plant staff and decision makers are properly trained in the 

use of the equipment. Station operators can obtain SFP level data trends and report those to 

decision makers within 30 minutes of request. The displays are inside a seismic structure that 

provides protection from adverse weather or flooding. 

In its letter dated August 28, 2014 [Reference 36], the licensee further stated that the backup 

display will be mounted in the Radwaste Control Panel Room Cabinet C104 at the 590 ft. 

elevation of the Auxiliary Building. This cabinet is located in Room 121, and can be accessed 

via Stairwell No. 16 and Door 75, or via Corridor 106A through Door 190. Both Stairwell 16 and 
- 77 -

Corridor 106A can be approached from the north, east, and west via Corridor 106. The back-up 

channel display can be considered promptly accessible, because it can be reached within the 

30 minute deployment requirement that exists for portable instrumentation. The impact to 

habitability would be primarily from elevated temperatures, as the C-40 panel room is 

considered a mild radiation environment. Personnel will not be continuously stationed at the 

backup display, it will be monitored periodically. The site FLEX Support Guidelines will provide 

guidance for personnel to evaluate the room temperature and take actions as necessary. 

The NRC staff noted that the licensee adequately addressed the display requirements. If 

implemented properly, the displays should provide continuous indication of SFP water level. 

The displays are located in seismically qualified buildings and the accessibility of the Radwaste 

Control Panel Room following an ELAP event is considered acceptable. The NRC staff finds 

that the licensee's proposed location and design of the SFP instrumentation displays appears to 

be consistent with NEI 12-02 guidance, as endorsed by JLD-ISG-2012-03, and should 

adequately address the requirements of the order. 

4.3 Evaluation of Programmatic Controls 

Order EA-12-051 specified that the spent fuel pool instrumentation shall be maintained available 

and reliable through appropriate development and implementation programmatic controls, 

including training, procedures, and testing and calibration. Below is the NRC staff's assessment 

of the programmatic controls for the spent fuel pool instrumentation. 

4.3.1 Programmatic Controls: Training 

In its OIP, the licensee stated that the Systematic Approach to Training (SAT) will be used to 

identify the population to be trained and to determine both the initial and continuing elements of 

the required training. Training will be completed prior to placing the instrumentation in service. 

The NRC staff finds that the use of SAT to identify the training population and to determine both 

the elements of the required training is acceptable. The licensee's proposed plan to train 

personnel in the operation, maintenance, calibration, and surveillance of the SFPI and the 

provision of alternate power to the primary and backup instrument channels, including the 

approach to identify the population to be trained, appears to be consistent with NEI 12-02 

guidance, as endorsed by JLD-ISG-2012-03, and should adequately address the requirements 

of the order. 

4.3.2 Programmatic Controls: Procedures 

In its OIP, the licensee stated that procedures for maintenance and testing will be developed 

using regulatory guidelines and vendor instructions. The BDBEE operation guidance will also 

address the following: 

• A strategy to ensure SFP water addition is initiated at an appropriate time consistent with 

implementation of NEI 12-06, Rev. 1. 

• Restoration of non-functioning SFP level channels after an event. Restoration timing will 

be consistent with the emergency condition. After an event, commercially available 

components that may not meet all qualifications may be used to replace components to 

restore functionality. 
- 78 -

During the onsite audit, the licensee further stated that Palisades SFP instrument channels 

automatically monitor the integrity of the measurement system using in-situ capability or onboard diagnostics. Deviation of measured test parameters from manufactured or as-installed 

configuration beyond a configurable threshold prompts operator intervention. Channel design 

provides capability for calibration or validation against known/actual SFP level. The Palisades 

SFP instrument channels have a reasonably high certified design accuracy of equal to or better 

than+/- 3 in. (excluding boric acid deposition effects that cause a conservative decrease in 

indicated level). Corrective Action CA-1 O has been initiated for condition report CR-PLP-2015-

2482 to track completion of the generation of a calibration procedure for the SFPI. 

Guideline FSG-11, "Alternate SFP Makeup and Cooling," provides actions to restore SFP level 

using an alternate makeup source for a BDBEE resulting in an ELAP. This procedure includes 

remote SFP level indicator display locations and a procedure for how and when to connect an 

external DC source to power the SFP level indicator. 

The NRC staff noted that the licensee adequately addressed the procedure requirements to 

establish procedures for the testing, surveillance, calibration, and operation of the primary and 

backup SFP level instrument channels. The NRC staff finds that the licensee's proposed 

procedures appear to be consistent with NEI 12 02, as endorsed by JLD-ISG-2012-03, and 

should adequately address the requirements of the order. 

4.3.3 Programmatic Controls: Testing and Calibration 

In its OIP, the licensee stated that procedures and PM tasks will be developed to perform 

required surveillance testing, calibration, backup battery maintenance, functional checks, and 

visual inspections of the probes. 

In its letter dated August 19, 2013 [Reference 32], the licensee further stated that the SFPI 

maintenance and testing program requirements ensure design and system readiness are 

planned to be established in accordance with Entergy's processes and procedures and in 

consideration of vendor recommendations to ensure that appropriate regular testing, channel 

checks, functional tests, periodic calibration, and maintenance is performed. 

During the onsite audit, the NRC staff inquired about the specific compensatory actions for the 

for out-of-service SFPI channel(s). In response, licensee provided stated that control of 

compensatory actions for out of service SFPI channel(s) will be controlled by inclusion in the 

Plant's Operating Requirements Manual. Compensatory actions for extended out-of-service 

events are summarized as follows: 

CONDITION REQUIRED ACTION COMPLETION 

TIME 

1. The Primary OR Back-up 1.a Restore spent fuel pool level 90 days 

Spent Fuel Pool level instrument to FUNCTIONAL status 

instrument does not meet the 

FUNCTIONAL requirements 

2. The Primary AND 2.a Initiate actions to restore 24 hours 

Back-up Spent Fuel Pool one of the channels of 
- 79 -

level instruments do not instrumentation to 

meet FUNCTIONAL FUNCTIONAL status. 

requirements. AND 

2.b Implement compensatory 72 hours 

measures such as use of 

alternate suitable 

equipment or supplemental 

personnel. 

3. Required Action and 3.a Implement compensatory Immediately 

associated Completion measures such as use of 

Time of Condition 1 not met. alternate suitable 

equipment or supplemental 

personnel. 

During the onsite audit, the licensee also provided a list of PM tasks planned to be developed 

for the Entergy fleet is as follows: 

Task Name Objective Frequency of 

Occurrence 

Channel Calibration Check To validate that the Mohr Seven Days 

(Operator Rounds) instruments (both channels) are 

eSOMS Tour displaying the correct spent fuel 

pool level within the accuracy of the 

instruments and that the date 

stamp on the display is indicating 

correctly. 

Channel Check/Panel To check each channel against 1 Year 

Functional Check each other for comparison and to 

perform functional assessments of 

each panel. 

Signal Processor Clock Battery To prevent failure of the onboard 10 Year 

Replacement clock battery and adverse impact to 

the signal processor operating 

system. 

Flush Probe Head Assembly Based upon boron precipitation As necessary via 

causing fouling of level detector corrective action. 

probes. 
- 80 -

The NRG staff noted that the licensee adequately addressed necessary testing and calibration 

of the primary and backup SFP level instrument channels to maintain the instrument channels at 

the design accuracy. The testing and calibration are consistent with the vendor 

recommendations. Additionally, compensatory actions for instrument channel(s) out-of-service 

appear to be consistent with guidance in NEI 12-02. 

The NRG staff finds that the licensee's proposed testing and calibration plan appears to be 

consistent with NEI 12 02, as endorsed by JLD-ISG-2012-03, and should adequately address 

the requirements of the order. 

4.4 Conclusions for Order EA-12-051 

In its letter dated December 16, 2015 [Reference 39], the licensee stated that they would meet 

the requirements of Order EA-12-051 by following the guidelines of NEI 12-02, as endorsed by 

JLD-ISG-2012-03. In the evaluation above, the NRG staff finds that, if implemented 

appropriately, the licensee has conformed to the guidance in NEI 12-02, as endorsed by JLDISG-2012-03. In addition, the NRG staff concludes that if the SFP level instrumentation is 

installed at Palisades Nuclear Station according to the licensee's proposed design, it should 

adequately address the requirements of Order EA-12-051. 

5.0 CONCLUSION 

In August 2013 the NRG staff started audits of the licensee's progress on Orders EA-12-049 

and EA-12-051. The staff conducted an onsite audit in June, 2015 [Reference 18]. The 

licensee reached its final compliance date on October 19, 2015, and has declared that the 

reactor is in compliance with the orders. The purpose of this safety evaluation is to document 

the strategies and implementation features that the licensee has committed to. Based on the 

evaluations above, the NRG staff concludes that the licensee has developed guidance and 

proposed designs that if implemented appropriately should adequately address the 

requirements of Orders EA-12-049 and EA-12-051. The NRG staff will conduct an onsite 

inspection to verify that the licensee has implemented the strategies and equipment to 

demonstrate compliance with the orders. 
- 81 -

6.0 REFERENCES 

1. SECY-11-0093, "Recommendations for Enhancing Reactor Safety in the 2P1 Century, 

the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," 

July 12, 2011 (ADAMS Accession No. ML 11186A950) 

2. SECY-12-0025, "Proposed Orders and Requests for Information in Response to 

Lessons Learned from Japan's March 11, 2011, Great Tohoku Earthquake and 

Tsunami," February 17, 2012 (ADAMS Accession No. ML 12039A 103) 

3. SRM-SECY-12-0025, "Staff Requirements - SECY-12-0025 - Proposed Orders and 

Requests for Information in Response to Lessons Learned from Japan's March 11, 

2011, Great Tohoku Earthquake and Tsunami," March 9, 2012 (ADAMS Accession No. 

ML 120690347) 

4. Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events," March 12, 2012 (ADAMS 

Accession No. ML 12054A736) 

5. Order EA-12-051, "Order Modifying Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation," March 12, 2012 (ADAMS Accession No. ML 12054A679) 

6. Nuclear Energy Institute document NEI 12-06, "Diverse and Flexible Coping Strategies 

(FLEX) Implementation Guide," Rev. 0, August 21, 2012 (ADAMS Accession No. 

ML 12242A378) 

7. JLD-ISG-2012-01, "Compliance with Order EA-12-049, Order Modifying Licenses with 

Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External 

Events," August 29, 2012 (ADAMS Accession No. ML 12229A 174) 

8. Nuclear Energy Institute document NEI 12-02, "Industry Guidance for Compliance with 

NRG Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation," Rev. 1, dated August 24, 2012 (ADAMS Accession No. ML 12240A307) 

9. JLD-ISG-2012-03, "Compliance with Order EA-12-051, Order Modifying Licenses with 

Regard to Reliable Spent Fuel Pool Instrumentation," August 29, 2012 (ADAMS 

Accession No. ML 12221 A339) 

1 O. Palisades Overall Integrated Plan in Response to March 12, 2012 Commission Order to 

Modify Licenses With Regard to Requirements for Mitigation Strategies for BeyondDesign-Basis External Events (Order Number EA-12-049) dated February 28, 2013 

(ADAMS Accession No. ML 13246A399) 

11. Palisades Nuclear Plant First Six-Month Status Report in Response to March 12, 2012 

Commission Order Modifying Licenses with Regard to Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) dated 

August 28, 2013 (ADAMS Accession No. ML13241A234) 
- 82 -

12. Palisades Nuclear Plant Second Six-Month Status Report in Response to March 12, 

2012 Commission Order Modifying Licenses with Regard Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) dated 

February 28, 2014 (ADAMS Accession No. ML 14059A078) 

13. Palisades Nuclear Plant Third Six-Month Status Report in Response to March 12, 2012 

Commission Order Modifying Licenses with Regard Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) dated 

August 28, 2014 (ADAMS Accession No. ML 14240A279) 

14. Palisades Nuclear Plant Fourth Six-Month Status Report in Response to March 12, 2012 

Commission Order Modifying Licenses with Regard Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) dated 

February 27, 2015 (ADAMS Accession No. ML 15062A011) 

15. Palisades Nuclear Plant Fifth Six-Month Status Report in Response to March 12, 2012 

Commission Order Modifying Licenses with Regard Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) dated 

August 28, 2015 (ADAMS Accession No. ML 15240A074) 

16. Letter from Jack R. Davis (NRG) to All Operating Reactor Licensees and Holders of 

Construction Permits, "Nuclear Regulatory Commission Audits of Licensee Responses 

to Mitigation Strategies Order EA-12-049," August 28, 2013 (ADAMS Accession No. 

ML 13234A503) 

17. Palisades Nuclear Plant - Interim Staff Evaluation Regarding Overall Integrated Plan in 

Response to Order EA-12-049 (Mitigation Strategies) (TAC NO. MF0768) dated 

February 10, 2014 (ADAMS Accession No. ML 13365A264) 

18. NRG Letter to Entergy, Palisades Nuclear Plant - Report for the Audit Regarding 

Implementation of Mitigating Strategies and Reliable Spent Fuel Pool Instrumentation 

Related of Orders EA-12-049 and EA-12-051 dated October 13, 2015 (ADAMS 

Accession No. ML 15272A324) 

19. Entergy Letter, Notification of Full Compliance with NRG Order EA-12-049, "Order 

Modifying Licenses with Regard to Requirements for Mitigation Strategies for BeyondDesign-Basis External Events,'' dated December 16, 2015 (ADAMS Accession No. 

ML 15351 A369) and Final Integrated Plan, dated December 13, 2015 (ADAMS 

Accession No. ML 15351 A360) 

20. U.S. Nuclear Regulatory Commission, "Request for Information Pursuant to Title 1 O of 

the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 

9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi 

Accident," March 12, 2012 (ADAMS Accession No. ML 12053A340) 

21. SRM-COMSECY-14-0037, "Staff Requirements - COMSECY-14-0037 - Integration of 

Mitigating Strategies For Beyond-Design-Basis External Events and the Reevaluation of 

Flooding Hazards," March 30, 2015 (ADAMS Accession No. ML 15089A236) 
- 83 -

22. Letter from Jack R. Davis (NRG) to Joseph E. Pollock (NEI), "Staff Assessment of 

National SAFER Response Centers Established In Response to Order EA-12-049," 

September 26, 2014 (ADAMS Accession No. ML 14265A 107) 

23. NRG Office of Nuclear Reactor Regulation Office Instruction LIC-111, Regulatory Audits, 

dated December 16, 2008 (ADAMS Accession No. ML082900195) 

24. Letter from William Dean (NRG) to Power Reactor Licensees, "Coordination of Request 

for Flooding Hazard Reevaluations and Mitigating Strategies for Beyond-Design Bases 

External Events," dated September 1, 2015 (ADAMS Accession No. ML 15174A257) 

25. NEI Position Paper: "Shutdown/Refueling Modes," dated September 18, 2013 (ADAMS 

Accession No. ML 13273A514) 

26. Letter from Jack R. Davis (NRG) to Joseph E. Pollock (NEI), regarding NRG 

endorsement of NEI Position Paper: "Shutdown/Refueling Modes," dated September 30, 

2013 (ADAMS Accession No. ML 13267 A382) 

27. COMSECY-14-0037, "Integration of Mitigating Strategies for Beyond-Design-Basis 

External Events and the Reevaluation of Flooding Hazards," dated November 21, 2014 

(ADAMS Accession No. ML 14309A256) 

28. Nuclear Energy Institute document NEI 12-06, "Diverse and Flexible Coping Strategies 

(FLEX) Implementation Guide," Rev. 2, December 31, 2015 (ADAMS Accession No. 

ML 16005A625) 

29. JLD-ISG-2012-01, "Compliance with Order EA-12-049, Order Modifying Licenses with 

Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External 

Events," Revision 1, January 22, 2012 (ADAMS Accession No. ML 15357A163) 

30. Palisades Overall Integrated Plan in Response to March 12, 2012 Commission Order 

Modifying License With Regard to Reliable Spent Fuel Pool Instrumentation (Order 

Number EA-12-051) dated February 28, 2013 (ADAMS Accession No. ML 13060A360) 

31. NRG Request For Additional Information Regarding Spent Fuel Pool Instrumentation 

(Order EA-12-051), July 18, 2013 (ADAMS Accession No. ML13200A328) 

32. Palisades Nuclear Plant, Response to Request for Additional Information for the Overall 

Integrated Plan in Response to the Commission Modifying Licenses with Regard to 

Requirements for Reliable Spent Fuel Pool Instrumentation (Order EA-12-051 ), dated 

August 19, 2013 (ADAMS Accession No. ML 13231A126) 

33. NRC Issuance of Palisades Interim Staff Evaluation and Request For Additional 

Information Regarding The Overall Integrated Plan For Implementation Of Order EA-12-

051, Reliable Spent Fuel Pool Instrumentation, dated November 26, 2013 (ADAMS 

Accession No. ML 13312A423). 
- 84 -

34. Palisades Nuclear Plant First Six-Month Status Report in Response to March 12, 2012 

Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation (Order Number EA-12-051), dated August 28, 2013 (ADAMS Accession 

No. ML 13241 A235). 

35. Palisades Nuclear Plant-Second Six-Month Status Report in Response to March 12, 

2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation (Order Number EA-12-051), dated February 28, 2014 (ADAMS 

Accession No. ML 14059A077). 

36. Palisades, Third Six-Month Status Report in Response to March 12, 2012 Commission 

Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation 

(Order Number EA-12-051 ), dated August 28, 2014 (ADAMS Accession No. 

ML 14240A278). 

37. Palisades Nuclear Plant, Fourth Six-Month Status Report in Response to March 12, 

2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation (Order Number EA-12-051), dated February 27, 2015 (ADAMS 

Accession No. ML 15062A056). 

38. Palisades Nuclear Plant Fifth Six-Month Status Report in Response to March 12, 2012 

Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation (Order Number EA-12-051), dated August 28, 2015 (ADAMS Accession 

No. ML 15240A074). 

39. Palisades - Notification of Full Compliance with NRG Order EA-12-051, "Order to Modify 

Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", dated December 16, 

2015 (ADAMS Accession No. ML 15351A 126). 

40. U.S. NRG, "Donald C. Cook Nuclear Plant, Units 1 And 2- Report For The Onsite Audit 

of Mohr Regarding Implementation of Reliable Spent Fuel Pool Instrumentation Related 

To Order EA-12-051," August 27, 2014 (ADAMS Accession No. ML 14216A362) 

41. U.S. Nuclear Regulatory Commission, "Mitigation of Beyond-Design-Basis Events," 

Federal Register, Vol. 80, No. 219, November 13, 2015, pp. 70610-70647. 

42. Nuclear Energy Institute document NEI 12-06, "Diverse and Flexible Coping Strategies 

(FLEX) Implementation Guide," Revision 2, dated December 31, 2015 (ADAMS 

Accession No. ML 16005A625) 

43. JLD-ISG-2012-01, "Compliance with Order EA-12-049, Order Modifying Licenses with 

Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External 

Events," Revision 1, dated January 22, 2012 (ADAMS Accession No. ML 15357A163) 
Principal Contributors: 

Date: August 22, 2016 

A. Roberts 

J. Miller 

J. Lehning 

M. McConnell 

K. Nguyen 

B. Lee 

K. Roche 

J. Hughey 

- 85 -

D. Johnson (MTS) 
- 2 -

October 13, 2015 (ADAMS Accession No. ML 15272A324), the NRC staff issued an ISE and 

audit report, respectively, on the licensee's progress. By letter dated March 26, 2014 (ADAMS 

Accession No. ML 14083A620), the NRC notified all licensees and construction permit holders 

that the staff is conducting audits of their responses to Order EA-12-051 in accordance with 

NRC NRR Office Instruction LIC-111, similar to the process used for Order EA-12-049. By letter 

dated December 16, 2015 (ADAMS Accession No. ML 15351A126), Entergy submitted a 

compliance letter in response to Order EA-12-051. The compliance letter stated that the 

licensee had achieved full compliance with Order EA-12-051. 

The enclosed safety evaluation provides the results of the NRC staffs review of Entergy's 

strategies for Palisades. The intent of the safety evaluation is to inform Entergy on whether or 

not its integrated plans, if implemented as described, provide a reasonable path for compliance 

with Orders EA-12-049 and EA-12-051. The staff will evaluate implementation of the plans 

through inspection, using Temporary Instruction 191, "Implementation of Mitigation Strategies 

and Spent Fuel Pool Instrumentation Orders and Emergency Preparedness 

Communications/Staffing/ Multi-Unit Dose Assessment Plans" (ADAMS Accession No. 

ML 15257A188). This inspection will be conducted in accordance with the NRC's inspection 

schedule for the plant. 

If you have any questions, please contact John Hughey, Orders Management Branch, 

Palisades Project Manager, at 301-415-3204 or at John.Hughey@nrc.gov. 

Docket No.: 50-255 

Enclosure: 

Safety Evaluation 

cc w/encl: Distribution via Listserv 

DISTRIBUTION: 

PUBLIC 

JOMB R/F 

RidsNrrDorllpl 3-1 Resource 

RidsNrrPMPalisades Resource 

RidsNrrLASLent Resource 

JQuichoch, NRR/JLD 

ADAMS Accession No. ML16014A318 

OFFICE NRR/JLD/JOMB/PM NRR/JLD/LA 

NAME JHughey Slent 

DATE 08/02/2016 08/02/2016 

OFFICE NRR/JLD/JOMB/BC (A) 

NAME MHalter 

DATE 08/22/2016 

Sincerely, 

/RA/ 

Mandy K. Halter, Acting Chief 

Orders Management Branch 

Japan Lessons-Learned Division 

Office of Nuclear Reactor Regulation 

RidsRgn3MailCenter Resource 

JHughey, NRR/JLO 

MHalter, NRR/JLD 

RidsAcrsAcnw_MailCTR Resource 

SBailey, NRR/JLO 

*via e-mail 

NRR/JLO/JCBB/BC NRR/JLD/JERB/BC 

JQuichocho * SBailey * 

08/05/2016 08/09/2016 

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