ML18201A492: Difference between revisions

2. Inspection Report 05000382/2017009

Mr. John Dinelli, Site Vice President

Entergy Operations, Inc.

17265 River Road

Killona, LA 70057-0751

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - INSPECTION OF THE

IMPLEMENTATION OF MITIGATION STRATEGIES AND SPENT FUEL POOL

INSTRUMENTATION ORDERS AND EMERGENCY PREPAREDNESS

COMMUNICATION/STAFFING/MULTI-UNIT DOSE ASSESSMENT PLANS -

INSPECTION REPORT 05000382/2017009 AND NOTICE OF VIOLATION

Dear Mr. Dinelli:

On September 21, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed the

on-site portion of the subject inspection at your Waterford Steam Electric Station, Unit 3, and

continued in-office review and inspection through June 2018. On June 7, 2018, the NRC

inspectors discussed the final results of this inspection with you and other members of your

staff. The results of this inspection are documented in the enclosed report.

The inspection examined activities conducted under your license as they relate to the

implementation of mitigation strategies and spent fuel pool instrumentation orders (EA-12-049

and EA-12-051) and emergency preparedness communication, staffing, and multi-unit dose

assessment plans, your compliance with the Commissions rules and regulations, and with the

conditions of your operating license. Within these areas, the inspection involved examination of

selected procedures and records, observation of activities, and interviews with station

personnel.

The enclosed report discusses a violation associated with a finding of very low safety

significance (Green). The NRC evaluated this violation in accordance Section 2.3.2 of the

NRC Enforcement Policy, which can be found at http://www.nrc.gov/about-nrc/regulatory/

enforcement/enforce-pol.html. The violation met the requirements for treatment as a non-cited

violation; however, because the violation is associated with a failure to meet the requirements of

an Order issued by the Commission which will require subsequent, specific follow-up inspection

to ensure compliance has been established, the NRC determined the issuance of Notice of

Violation (Notice) is appropriate in this case.

You are required to respond to this letter and should follow the instructions specified in the

enclosed Notice when preparing your response. The NRCs review of your response will also

determine whether further enforcement action or inspection is necessary to ensure your

compliance with regulatory requirements.

J. Dinelli

2

Additionally, the NRC inspectors documented one finding of very low safety significance (Green)

in this report. The finding did not involve a violation of NRC requirements.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a

regulatory requirement in this report, you should provide a response within 30 days of the date

of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory

Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the

Regional Administrator, Region IV; and the NRC resident inspector at the Waterford Steam

Electric Station, Unit 3.

This letter, its enclosures, and your response will be made available for public inspection and

copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room

in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Geoffrey Miller, Branch Chief

Project Branch D

Division of Reactor Projects

Docket No. 50-382

License No. NPF-38

Enclosures:

1. Notice of Violation

w/Attachment

cc: Electronic copy for Waterford 3

Enclosure 1

NOTICE OF VIOLATION

Entergy Operations, Inc.

Docket No. 50-382

Waterford Steam Electric Station, Unit 3

License No. NPF-38

Enterprise Identifier: I-2017-009-0006

During an NRC inspection conducted from September 18, 2017, through June 7, 2018, a

violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy,

the violation is listed below:

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

Strategies for Beyond-Design-Basis External Events, dated March 12, 2012,

Section IV.A.2, requires, in part, that all licensees identified in Attachment 1 to this Order

comply with the requirements described in Attachment 2 of this Order except to the extent

that a more stringent requirement is set forth in the license.

Order EA-12-049, Attachment 1, identified Entergy Operations, Inc., Waterford Steam

Electric Station, Unit 3, (Waterford 3) as a power reactor licensee subject to Section IV of

the Order.

Order EA-12-049, Attachment 2, requires, in part, that licensees develop, implement, and

maintain guidance and strategies to maintain or restore core cooling, containment and spent

fuel pool cooling capabilities following a beyond-design-basis external event. The transition

phase requires providing sufficient, portable, on-site equipment and consumables to

maintain or restore these functions until they can be accomplished with resources brought

from off site. Licensees must also provide reasonable protection for the associated

equipment from external events, and full compliance includes, in part, the staging or

installation of equipment needed for the strategies.

ENTGWF081-REPT-001, Waterford Steam Electric Station Unit 3 Final Integrated Plan,

Revision 1, dated July 20, 2016, provides the necessary guidance on strategies to maintain

or restore core cooling, containment, and spent fuel pool cooling capabilities following a

beyond-design-basis external event.

Section 2.2, Strategies, states, in part, that Phase 2 strategies support the

transition from installed plant equipment to FLEX equipment which is deployed by

the on-shift personnel to maintain essential functions.

Section 2.3.2, Phase 2 Strategy, states, in part, that the capability for reactor core

cooling is accomplished from a pre-staged FLEX core cooling pump to provide

feedwater to the steam generators in the event the turbine-driven emergency

feedwater pump fails or sufficient steam pressure is no longer available to drive the

turbine-driven emergency feedwater pump turbine, and that the FLEX core cooling

pump is powered by the FLEX diesel generator. Section 2.3.2 also states, in part,

that reactor coolant system inventory control involves the use of refueling water

storage pool or boric acid makeup tank inventory through a repowered charging

pump which receives its power from the FLEX diesel generator. Section 2.3.2 further

states, in part, that the FLEX diesel generator is capable of supplying power to a

battery charger such that DC power for controls and instrumentation continues to be

available to support the reactor coolant system core cooling function.

2

Enclosure 1

Section 2.4.2, Phase 2 Strategy Modes 1-4, states in part, that the capability to

provide spent fuel pool make-up and/or spray during Phase 2 is accomplished using

the component cooling water make-up pumps which are powered by the FLEX diesel

generator.

Section 2.7, Planned Protection of FLEX Equipment, states, in part, that in order to

assure reliability and availability of the FLEX equipment required by the FLEX

strategy, Waterford 3 has sufficient equipment to address all functions on-site, plus

one additional spare (i.e., an N+1 capability). Section 2.7 further states, in part,

that the N+1 diesel generator provides the capability to restore the N function by

relocating the N+1 diesel generator to the reactor auxiliary building from the N+1

storage building.

Section 2.15.1, Method of Storage and Protection of FLEX Equipment, states, in

part, that to assist with unanticipated unavailability of the N set, evaluations have

been performed and pre-planned strategies have been developed to provide

reasonable protection of specific N+1 equipment for predictable external events

with pre-warning (i.e., Mississippi River flood and hurricanes) and instances where

the N set is unavailable for conditions other than conduct of routine maintenance

and testing during normal operations.

Contrary to the above, from June 1, 2016, to June 7, 2018, the licensee failed to adequately

develop, implement, and maintain guidance and strategies to maintain or restore core

cooling, containment and spent fuel pool cooling capabilities following a beyond-design-

basis external event. Specifically, the licensee failed to establish appropriate design and

procedures associated with providing electrical power using the N+1 FLEX diesel

generator to support transition phase (Phase 2) strategies necessary to maintain or restore

the core cooling and spent fuel pool cooling capabilities in mitigating a beyond-design-basis

external event.

This violation is associated with a Green Significance Determination Process finding.

Pursuant to the provisions of 10 CFR 2.201, Entergy Operations, Inc., is hereby required to

submit a written statement or explanation to the U.S. Nuclear Regulatory Commission,

ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional

Administrator, Region IV, 1600 E. Lamar Blvd., Arlington, TX 76011-4511, and a copy to the

NRC resident inspector at the facility that is the subject of this Notice of Violation (Notice), within

30 days of the date of the letter transmitting this Notice. This reply should be clearly marked as

a Reply to a Notice of Violation; EA-18-020 and should include for the violation: (1) the reason

for the violation or, if contested, the basis for disputing the violation or severity level, (2) the

corrective steps that have been taken and the results achieved, (3) the corrective steps that will

be taken, and (4) the date when full compliance will be achieved.

Your response may reference or include previous docketed correspondence, if the

correspondence adequately addresses the required response. If an adequate reply is not

received within the time specified in this Notice, an order or a Demand for Information may be

issued as to why the license should not be modified, suspended, or revoked, or why such other

action as may be proper should not be taken. Where good cause is shown, consideration will

be given to extending the response time.

3

Enclosure 1

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC

Public Document Room or from the NRCs Agencywide Documents Access and Management

System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-

rm/adams.html, to the extent possible, it should not include any personal privacy or proprietary

information so that it can be made available to the public without redaction. If personal privacy

or proprietary information is necessary to provide an acceptable response, then please provide

a bracketed copy of your response that identifies the information that should be protected and a

redacted copy of your response that deletes such information. If you request withholding of

such material, you must specifically identify the portions of your response that you seek to have

withheld and provide in detail the bases for your claim of withholding (e.g., explain why the

disclosure of information will create an unwarranted invasion of personal privacy or provide the

information required by 10 CFR 2.390(b) to support a request for withholding confidential

commercial or financial information).

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working

days of receipt.

Dated this 20th day of July 2018

Enclosure 2

U.S. NUCLEAR REGULATORY COMMISSION

Inspection Report

Docket Number(s):

05000382

License Number(s):

NPF-38

Report Number(s):

05000382/2017009

Licensee:

Entergy Operations, Inc.

Facility:

Waterford Steam Electric Station, Unit 3

Location:

Killona, Louisiana

Inspection Dates:

September 18, 2017 to June 7, 2018

Inspectors:

R. Alexander, Senior Project Engineer (Team Leader)

J. Mateychick, Senior Reactor Inspector

C. Speer, Resident Inspector

E. Uribe, Project Engineer

D. Loveless, Senior Reactor Analyst

Approved By:

G. Miller

Chief, Project Branch D

Division of Reactor Projects

2

Enclosure 2

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensees performance

by conducting a Temporary Instruction 2515/191, Implementation of Mitigation Strategies and

Spent Fuel Pool Instrumentation Orders and Emergency Preparedness Communication/

Staffing/Multi-Unit Dose Assessment Plans at Waterford 3 Steam Electric Station in

accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs

program for overseeing the safe operation of commercial nuclear power reactors. Refer to

https://www.nrc.gov/reactors/operating/oversight.html for more information. NRC and self-

revealed findings, violations, and additional items are summarized in the table below.

List of Findings and Violations

Failure to Establish Appropriate Electrical-Related FLEX Strategies for Mitigating a Beyond-

Design-Basis External Event

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Strategies

Green Notice of Violation

VIO 05000382/2017009-01

Open

H.14 -

Conservative

Bias

The NRC inspection team identified a Green finding and associated violation in that the

licensee failed to adequately develop, implement, and maintain guidance and strategies to

maintain or restore core cooling, containment and spent fuel pool cooling capabilities following

a beyond-design-basis external event. Specifically, the licensee failed to establish appropriate

design and procedures associated with providing electrical power using the N+1 FLEX diesel

generator to support transition phase (Phase 2) strategies necessary to maintain or restore the

core cooling and spent fuel pool cooling capabilities in mitigating a beyond-design-basis

external event.

Failure to Adequately Consider the Impacts on FLEX Phase 2 Equipment from Large Internal

Flooding Sources Which Were Not Seismically Robust

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green Finding

FIN 05000382/2017009-02

Closed

P.2 - Evaluation

spent fuel pool instrumentation.

The NRC inspection team identified a Green finding related to the licensees failure to

adequately consider the impacts from large internal flooding sources that are not seismically

robust on the design, protection, and staging of the FLEX core cooling pump on the -35 ft.

elevation of the reactor auxiliary building.

3

Enclosure 2

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in

effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with

their attached revision histories are located on the public website at http://www.nrc.gov/reading-

rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared

complete when the IP requirements most appropriate to the inspection activity were met

consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection

Program - Operations Phase. The team reviewed selected procedures and records, observed

activities, and interviewed personnel to assess licensee performance and compliance with

Commission rules and regulations, license conditions, site procedures, and standards.

OTHER ACTIVITIES - TEMPORARY INSTRUCTIONS

TI 2515/191 - Inspection of the Implementation of Mitigation Strategies and Spent Fuel Pool

Instrumentation Orders and Emergency Preparedness Communication/Staffing/Multi-Unit Dose

Assessment Plans

The inspection verified plans for complying with NRC Orders EA-12-049, Order Modifying

Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis

External Events (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML12056A045) and EA-12-051, Order Modifying Licenses With Regard to

Reliable Spent Fuel Pool Instrumentation (ADAMS No. ML12054A679) are in place and are

being implemented by the licensee. Additionally, the inspection verified implementation of

staffing and communications information provided in response to the March 12, 2012,

request for information letter (ADAMS No. ML12053A340) and multiunit dose assessment

information provided per COMSECY-13-0010, Schedule and Plans for Tier 2 Order on

Emergency Preparedness for Japan Lessons Learned, dated March 27, 2013,

(ADAMS No. ML12339A262).

(1) Based on samples selected for review, the team inspected to determine whether the

licensee satisfactorily implemented appropriate elements of the Diverse and Flexible

Coping Strategies (FLEX) as described in the plant specific submittals [including the

Final Integrated Plan (ADAMS No. ML16203A321)] and the associated safety evaluation

(ADAMS No. ML17045A148), and to determine whether the licensee is in compliance

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

Mitigation Strategies for Beyond-Design-Basis External Events. Specifically, the team

inspected to determine whether the licensee satisfactorily:

a) developed and issued FLEX Support Guidelines (FSGs) to implement the FLEX

strategies for postulated external events;

b) integrated their FSGs into their existing plant procedures such that entry into and

departure from the FSGs were clear when using existing plant procedures;

c) protected FLEX equipment from site-specific hazards;

d) developed and implemented adequate testing and maintenance of FLEX equipment

to ensure their availability and capability;

4

Enclosure 2

e) trained their staff to assure personnel proficiency in the mitigation of

beyond-design basis events; and

f) developed the means to ensure the necessary off-site FLEX equipment would be

available from off-site locations.

(2) Based on samples selected for review, the team inspected to determine whether the

licensee satisfactorily implemented appropriate elements of the FLEX strategy as

described in the plant specific submittals [including the Final Integrated Plan (ADAMS

No. ML16203A321)] and the associated safety evaluation (ADAMS No. ML17045A148),

and to determine whether the licensee is in compliance with NRC Order EA-12-051,

Order Modifying Licenses With Regard to Reliable Spent Fuel Pool Instrumentation.

Specifically, the team inspected to determine whether the licensee satisfactorily:

a) installed the spent fuel pool instrumentation sensors, cabling and power supplies to

provide physical and electrical separation as described in the plant specific

submittals and safety evaluation;

b) installed the spent fuel pool instrumentation display in the location, environmental

conditions, and accessibility as described in the plant specific submittals;

c) trained their staff to assure personnel proficiency with the maintenance, testing, and

use of the spent fuel pool instrumentation; and

d) developed and issued procedures for maintenance, testing, and use of the reliable

(3) The team reviewed information provided in the licensees multi-unit dose submittal

and in response to the NRCs March 12, 2012, request for information letter (ADAMS

No. ML12053A340), and inspected to determine whether the licensee satisfactorily

implemented enhancements pertaining to Near-Term Task Force Recommendation 9.3.

Specifically, the team inspected to determine whether:

a) the licensee satisfactorily implemented required staffing changes to support an

extended loss of all ac power (ELAP)/loss of ultimate heat sink (LUHS) scenario;

b) emergency preparedness communications equipment and facilities were sufficient

for dealing with an ELAP/LUHS scenario; and

c) the licensee implemented multi-unit/-source dose assessment capabilities (including

releases from the spent fuel pool) using the licensees site-specific dose assessment

software and approach.

The team verified that non-compliances with requirements and standards identified during the

inspection were entered into the licensee's corrective action program as appropriate.

5

Enclosure 2

INSPECTION RESULTS

Failure to Establish Appropriate Electrical-Related FLEX Strategies for Mitigating a Beyond-

Design-Basis External Event

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Strategies

Green Notice of Violation

VIO 05000382/2017009-01

Open

H.14 -

Conservative

Bias

The NRC inspection team identified a Green finding and associated violation in that the

licensee failed to adequately develop, implement, and maintain guidance and strategies to

maintain or restore core cooling, containment and spent fuel pool cooling capabilities

following a beyond-design-basis external event. Specifically, the licensee failed to establish

appropriate design and procedures associated with providing electrical power using the N+1

FLEX diesel generator to support transition phase (Phase 2) strategies necessary to maintain

or restore the core cooling and spent fuel pool cooling capabilities in mitigating a beyond-

design-basis external event.

Description: To meet the requirements of NRC Order EA-12-049, the licensee committed in

its Final Integrated Plan (FIP) to meet the requirements contained in NEI 12-06, Diverse and

Flexible Coping Strategies (FLEX) Implementation Guide, Revision 0, as endorsed by the

NRC. As described in the NEI 12-06 guidance, Section 3.2.2, in order to assure reliability

and availability of the FLEX equipment required to meet these capabilities [in responding to

an ELAP/LUHS event], 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. As such, for the single unit Waterford 3 site, the N (or N-set) equipment was

generally designated as the primary or minimum FLEX equipment for a given function or set

of functions, and the N+1 equipment is the one additional backup piece of equipment for the

same functions/set of functions.

However, the licensee requested alternatives to the NEI 12-06, Revision 0, guidance, in part,

because the Waterford 3 site location along the banks of the Mississippi River raised potential

challenges to moving FLEX equipment from the N+1 storage building during postulated

flooding or hurricane events. The alternatives, in part, included:

(1) the pre-staging of one 480 VAC FLEX diesel generator (the N FLEX DG) within a

robust structure inside the Nuclear Plant Island Structure (NPIS) on the reactor auxiliary

building roof at the +41 foot MSL elevation; and

(2) storage of some N+1 FLEX equipment (including the N+1 FLEX DG) in the N+1

storage building located outside of the NPIS, though the N+1 storage building is not

protected against flooding or wind driven missile hazards.

Additionally, given that the N+1 storage building was not designed to protect the backup

equipment stored within it from all hazards applicable to the station, the licensee proposed

that should any of the pre-staged N-set equipment inside the NPIS (including the N FLEX

DG) be out-of-service or non-functional, the licensee would initiate actions within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

and implement compensatory actions within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the N capability of that

equipment. These actions included providing protection for the N+1 equipment (e.g., N+1

6

Enclosure 2

FLEX DG) by moving them to within the NPIS within the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed to implement

compensatory measures. The NRC staff documented in its Safety Evaluation

(Sections 3.14.4 and 3.14.5) that although the guidance of NEI 12-06 has not been met, if

these alternatives are implemented as described by the licensee, they will meet the

requirements of the order.

The licensees FLEX electrical strategy (as described in the FIP Section 2.3.2 and 2.3.11)

includes shedding of non-critical DC electrical loads early in the event to allow the stations

DC battery coping time to be extended from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to approximately 12.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Upon

connection and starting of a FLEX DG, the station batteries could then be recharged providing

continuous DC power to critical instruments and valves throughout Phase 2 of the event.

Further, in addition to repowering the DC battery chargers, the FLEX DG was designed to

provide or restore electrical power to the following Phase 2 equipment: (1) permanently

installed charging pumps (to provide reactor coolant system makeup capability), (2) FLEX

core cooling pumps (which are backups to the turbine-driven emergency feedwater pump to

supply steam generator makeup capability), (3) permanently installed component cooling

water makeup pumps (to provide spent fuel pool makeup capability), (4) fuel oil transfer

pump, (5) battery room exhaust fan, and (6) various lighting.

During the course of the on-site inspection, the team identified that the licensee had

developed procedures during final implementation which were inconsistent with the FIP and

the NRC Safety Evaluation and failed to ensure that the key FLEX safety functions continued

to be met. Specifically:

(1) The team determined that the licensee procedures and guidelines were implemented

such that the N+1 FLEX DG would only be moved inside the NPIS (and protected) for the

concurrent conditions of both the N FLEX DG out-of-service and when the station was

forecast to experience predictable external events with pre-warning (e.g., hurricane,

Mississippi River flooding). The team determined that this was contrary to the approved

alternative described in the NRC Safety Evaluation in which the NRC staff understood that the

licensee would move the N+1 FLEX DG inside the NPIS when the N FLEX DG was out of

service or when the station was forecast to experience predictable external events to ensure

that the N FLEX DG capability was continuously met under all hazards.

(2) Additionally, the team determined that if the licensee were to move the N+1 FLEX DG

inside the NPIS for the conditions described above, the licensee had not established

procedures or provisions to be able to run the N+1 FLEX DG inside the NPIS as described by

NEI 12-06, Revision 0, Sections 6.2.3 and 7.3.2. Specifically, the licensee had not

established a means to appropriately fuel or ventilate the N+1 FLEX DG had it become

necessary to operate the engine inside of the enclosed reactor auxiliary building during a

persistent flood or hurricane. Therefore the capability of the FLEX DGs could not be met

under all hazards by the N+1 FLEX DG if the N FLEX DG was non-functional.

(3) Finally, the team determined that the licensee implemented part of the provisions of

NEI 12-06, Revision 2, Section 11.5, which allows an outage time of 45 days for the N FLEX

DG when the N+1 FLEX DG was in a location that was not fully protected under all hazards.

The team determined that the licensee implemented this revised allowed outage time for a

25-day period beginning in February 2017 when the N FLEX DG was out of service due to a

failed component when there was not a concurrent predicable external event, and therefore

7

Enclosure 2

the licensee staged the N+1 FLEX DG outside the NPIS at the ground elevation of the station

unprotected from potential high winds, hurricanes, seismic, and/or flooding hazards.

The inspection team further determined that in implementing the approved alternatives

described in the NRC Safety Evaluation for the protection of the N+1 FLEX DG when the

N FLEX DG was out of service for other than planned maintenance, the licensee had created

a Technical Requirements Manual (TRM) Limiting Condition for Operation Action Statement

that was inconsistent with the approved alternatives. Specifically, TRM Limiting Condition for

Operation 3.13.2, Action b.1, correctly implemented the approved alternatives in the Safety

Evaluation and NEI 12-06, Revision 0, stating:

With one or more N RAB [Reactor Auxiliary Building] FLEX Components specified in

Table 3.13-1 [which includes the FLEX DG] relocate the N+1 Building FLEX

component to meet N RAB FLEX capability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and restore the N RAB

FLEX component within 90 days if the N+1 Building FLEX component relocation can

satisfy the FLEX functions for all events.

However, TRM Limiting Condition for Operation 3.13.2, Action b.2, went further by

implementing elements of NEI 12-06, Revision 2, in stating:

With one or more N RAB FLEX Components specified in Table 3.13-1 relocate the

N+1 Building FLEX component to meet N RAB FLEX capability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and

restore the N RAB FLEX component within 45 days if the N+1 Building FLEX

component relocation cannot completely satisfy the FLEX functions for all events.

The 45-day allowed outage time as described in NEI 12-06, Revision 2, was intended for

situations where FLEX equipment could be pre-staged as risk reduction measures in

locations not fully protected from all hazards to facilitate shorter FLEX implementation

timelines when the unit was in Modes 5 or 6 (i.e., during refueling outages). In implementing

the reduced allowed outage time described in Revision 2, the N-set FLEX capability was still

required to be met, which was not the situation for the 25-day period for the N+1 FLEX DG as

identified by the inspection team. However, the licensees FIP clearly stated that the

licensees strategies were developed based on the guidance in NEI 12-06, Revision 0, and

the NRC Safety Evaluation stated that the staff evaluated the strategies against that standard

(Section 3.1, page 8).

Because the licensee committed to NEI 12-06, Revision 0, which does not have the revised

allowed outage time for the unprotected equipment, the inspection team determined that the

allowed outage time for the N FLEX DG should have been 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with the N+1 FLEX DG

staged in the location outside of the NPIS. As such, the licensee failed to ensure the N

capability was met for the FLEX DGs for a period of at least 25 days in early 2017, until the

failed component was replaced and the N FLEX DG returned to a functional and available

status.

The inspection team determined that the issues described above were examples associated

with the licensees overall failure to develop, implement, and maintain guidance and

strategies to maintain or restore core cooling, containment and spent fuel pool cooling

capabilities following a beyond-design-basis external event as required by NRC

Order EA-12-049.

8

Enclosure 2

Corrective Actions: Following teleconferences with the Region IV-based inspection team and

the NRC Headquarters Beyond-Design-Basis events subject matter experts in October

through November 2017, the licensee acknowledged the concerns raised by the inspection

team and documented in the corrective action program that the strategies to utilize N+1 FLEX

DG were not consistent with the guidance in NEI 12-06, Revision 0, nor the approved

alternatives. Additionally on November 16, 2017, the licensee used the correct 90-day action

statement in TRM Limiting Condition for Operation 3.13.2 for the N+1 FLEX DG not being

available under all conditions with the procedures currently in place for the station, while the

licensee continued to take actions to correct the issues identified by the inspection team.

Subsequently, on January 10, 2018, the licensee declared the N+1 FLEX DG functional (and

exited the Technical Requirements Manual Action Statement), with additional administrative

controls in place should the N FLEX DG become non-functional, while the licensee pursues

long term corrective actions to revise the strategy for the protection and utilization of the N+1

FLEX DG under all hazards.

Corrective Action Reference(s): CR-WF3-2017-09150; CR-WF3-2017-09152

Performance Assessment:

Performance Deficiency: The inspection team identified that the licensee established their

FLEX strategies for the protection and utilization of the N+1 FLEX Diesel Generator (DG)

which were inconsistent with the requirements of NEI 12-06, Revision 0, and the approved

alternatives to the guidance described in Section 12.5 of the licensees Final Integrated Plan,

and Section 3.14 of the NRC Safety Evaluation, which was a performance deficiency.

Specifically, the licensee established the FLEX strategies such that the N+1 FLEX DG could

not be used under all hazards, and implemented a revised allowed out-of-service time

inconsistent with the NEI 12-06, Revision 0, and approved alternative guidance.

Screening: The performance deficiency is more than minor because it is associated with the

equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects

its objective to ensure the availability, reliability, and capability of systems that respond to

initiating events to prevent undesirable consequences (i.e., core damage). Specifically, in

establishing the FLEX electrical strategy the licensee failed to ensure the N capability of the

FLEX DGs remained available under all hazards.

Significance: The team assessed the significance of the finding using IMC 0609, Appendix O,

Significance Determination Process for Mitigating Strategies and Spent Fuel Pool

Instrumentation (Orders EA-12-049 and EA-12-051), dated October 7, 2016. The team

determined that the performance deficiency did not impact the spent fuel pool instrumentation

order (EA-12-051) (question 1). However, the team determined that the performance

deficiency included an exposure period of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and involved both

unavailable equipment and deficient procedures associated with the N+1 FLEX DG that

would result in the complete loss of one or more FLEX functions (questions 2 and 3).

Additionally, using available information, the product of the findings exposure time and the

applicable external event initiating event frequencies was greater than the 1E-6 threshold [in

this case the hurricane event was the most dominant contributor] (question 4). Further, the

team determined that the performance deficiency could involve significant programmatic

issues in the areas of equipment/strategy design, procedural guidance, and training such that

the effectiveness of the Mitigating Strategy program is reduced (question 5). Therefore,

9

Enclosure 2

further evaluation of the finding was directed in accordance with IMC 0609, Appendix M,

Significance Determination Process Using Qualitative Criteria.

In accordance with IMC 0609, Appendix M, the senior reactor analyst completed a bounding

quantitative evaluation which provided an upper bound incremental conditional core damage

probability (ICCDP) of 1.32E-5, where the ICCDP contribution is dominated by a hurricane-

induced external event, with an exposure period capped at one year (from date of Order

compliance [May 31, 2016] through the date in which the inspectors identified the issue

[September 21, 2017]). The evaluation used a frequency of 4.6E-2/year for a hurricane-

initiated loss of offsite power (LOOP). This was the best estimate based on actual data given

that Waterford has experienced one hurricane-induced LOOP during the plant life (i.e., during

Hurricane Katrina, 2005).

Additionally, FLEX equipment and strategies required by Order EA-12-049 provide a level of

defense-in-depth for beyond-design-basis events resulting in an ELAP/LUHS in a similar

manner to the equipment and strategies required by 10 CFR 50.54(hh)(2) which are intended

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

under the circumstances associated with loss of large areas of the plant due to explosions or

fire. As such, the inspectors reviewed the guidance in IMC 0609, Appendix L, B.5.b

Significance Determination Process, for an additional qualitative bounding assessment for

this performance deficiency. In that the licensee failed to ensure the N-set capability of the

FLEX DG was maintained under all hazards, multiple FLEX strategies (i.e., core cooling

and spent fuel pool cooling) could not be assured under all hazards. Therefore, analogous

to the unavailable, unrecoverable loss of multiple functions intended to mitigate

beyond-design-basis events described in the IMC 0609, Appendix L, this FLEX-related

performance deficiency involved the loss of multiple functions intended to mitigate

beyond-design-basis events. As such, this deterministically developed methodology provided

a qualitative bounding assessment of low-to-moderate safety significance.

Finally, the NRC considered additional qualitative factors associated with (1) effects on

defense-in-depth; (2) extent the performance deficiency affected other equipment; (3) degree

of degradation of the equipment failure; (4) effects of exposure time on the performance

deficiency; (5) likelihood of the licensees recovery actions to mitigate the performance

deficiency; and (6) other qualitative circumstances. As further described in the attachment to

this report, the qualitative factors considered together with the bounding evaluations

described above support the final determination that the finding is of very low safety

significance (Green).

Cross-cutting Aspect: The finding had a human performance cross-cutting aspect associated

with conservative bias, in that the licensee failed to use a decision-making process that

emphasized prudent choices over those that were simply allowable. Specifically, the licensee

established the FLEX electrical strategy based in part on a review of the changes in

NEI 12-06, Revision 2, which appeared to allow the licensee to establish alternative controls

and procedures for the N+1 FLEX DG. However, in establishing this element of the strategy,

the licensee failed to consider the core requirement to ensure the N+1 FLEX DG would be

capable of providing the N-set capability under all hazards [H.14].

10

Enclosure 2

Enforcement:

Violation: Order EA-12-049, Order Modifying Licenses with Regard to Requirements for

Mitigating Strategies for Beyond-Design-Basis External Events, dated March 12, 2012,

Section IV.A.2, requires, in part, that all licensees identified in Attachment 1 to this Order

comply with the requirements described in Attachment 2 of this Order except to the extent

that a more stringent requirement is set forth in the license.

Order EA-12-049, Attachment 1, identified Entergy Operations, Inc., Waterford Steam Electric

Station, Unit 3, (Waterford 3) as a power reactor licensee subject to Section IV of the Order.

Order EA-12-049, Attachment 2, requires, in part, that licensees develop, implement, and

maintain guidance and strategies to maintain or restore core cooling, containment and spent

fuel pool cooling capabilities following a beyond-design-basis external event. The transition

phase requires providing sufficient, portable, on-site equipment and consumables to maintain

or restore these functions until they can be accomplished with resources brought from off site.

Licensees must also provide reasonable protection for the associated equipment from

external events and full compliance includes, in part, the staging or installation of equipment

needed for the strategies.

ENTGWF081-REPT-001, Waterford Steam Electric Station Unit 3 Final Integrated Plan,

Revision 1, dated July 20, 2016, provides the necessary guidance on strategies to maintain

or restore core cooling, containment, and spent fuel pool cooling capabilities following a

beyond-design-basis external event.

Section 2.2, Strategies, states, in part, that Phase 2 strategies support the transition

from installed plant equipment to FLEX equipment which is deployed by the on-shift

personnel to maintain essential functions.

Section 2.3.2, Phase 2 Strategy, states, in part, that the capability for reactor core

cooling is accomplished from a pre-staged FLEX core cooling pump to provide

feedwater to the steam generators in the event the turbine-driven emergency

feedwater pump fails or sufficient steam pressure is no longer available to drive the

turbine-driven emergency feedwater pump turbine, and that the FLEX core cooling

pump is powered by the FLEX diesel generator. Section 2.3.2 also states, in part, that

reactor coolant system inventory control involves the use of refueling water storage

pool or boric acid makeup tank inventory through a repowered charging pump which

receives its power from the FLEX diesel generator. Section 2.3.2 further states, in

part, that the FLEX diesel generator is capable of supplying power to a battery charger

such that DC power for controls and instrumentation continues to be available to

support the reactor coolant system core cooling function.

Section 2.4.2, Phase 2 Strategy Modes 1-4, states in part, that the capability to

provide spent fuel pool make-up and/or spray during Phase 2 is accomplished using

the component cooling water make-up pumps which are powered by the FLEX diesel

generator.

Section 2.7, Planned Protection of FLEX Equipment, states, in part, that in order to

assure reliability and availability of the FLEX equipment required by the FLEX

strategy, Waterford 3 has sufficient equipment to address all functions on-site, plus

11

Enclosure 2

one additional spare (i.e., an N+1 capability). Section 2.7 further states, in part, that

the N+1 diesel generator provides the capability to restore the N function by

relocating the N+1 diesel generator to the reactor auxiliary building from the N+1

storage building.

Section 2.15.1, Method of Storage and Protection of FLEX Equipment, states, in

part, that to assist with unanticipated unavailability of the N set, evaluations have

been performed and pre-planned strategies have been developed to provide

reasonable protection of specific N+1 equipment for predictable external events with

pre-warning (i.e., Mississippi River flood and hurricanes) and instances where the N

set is unavailable for conditions other than conduct of routine maintenance and testing

during normal operations.

Contrary to the above, from June 1, 2016, to June 7, 2018, the licensee failed to adequately

develop, implement, and maintain guidance and strategies to maintain or restore core cooling,

containment and spent fuel pool cooling capabilities following a beyond-design-basis external

event. Specifically, the licensee failed to establish appropriate design and procedures

associated with providing electrical power using the N+1 FLEX diesel generator to support

transition phase (Phase 2) strategies necessary to maintain or restore the core cooling and

spent fuel pool cooling capabilities in mitigating a beyond-design-basis external event.

Disposition: This violation met the requirements for treatment as a non-cited violation;

however, since the violation is associated with a failure to meet the requirements of orders

issued by the Commission which will require subsequent specific follow up inspection to

ensure compliance has been established, the NRC determined the issuance of Notice of

Violation is appropriate.

Failure to Adequately Consider the Impacts on FLEX Phase 2 Equipment from Large Internal

Flooding Sources Which Were Not Seismically Robust

Cornerstone

Significance

Cross-cutting

Aspect

Report

Section

Mitigating

Systems

Green Finding

FIN 05000382/2017009-02

Closed

P.2 - Evaluation

The NRC inspection team identified a Green finding related to the licensees failure to

adequately consider the impacts from large internal flooding sources that are not seismically

robust on the design, protection, and staging of the FLEX core cooling pump on the -35 ft.

elevation of the reactor auxiliary building.

Description: As described in the licensees FIP Section 2.3, the stations FLEX strategy for

core cooling utilizes the turbine driven emergency feedwater pump during Phase 1 of an

extended loss of ac power/loss of ultimate heat sink (ELAP/LUHS) event response (starting in

Modes 1-4) to feed the steam generators with feedwater from the condensate storage pool

and exhausting the steam via the atmospheric dump valves. In Phase 2 of the response, the

stations strategy continues to employ the turbine driven emergency feedwater pump since

sufficient steam pressure is expected to be available to drive the pump turbine, drawing

feedwater from the condensate storage pool and subsequently from the wet cooling tower

basins. However, as backup should the pump fail or steam pressure drop below that which is

necessary to drive the pump turbine, the licensee pre-staged an electric driven FLEX core

12

Enclosure 2

cooling pump (FCCP) at the reactor auxiliary building -35 ft. elevation. This pre-staged

backup pump is the N-set FCCP, while the N+1 FCCP is stored in the N+1 storage building.

Either FCCP can be powered the FLEX power panel connected to either the N or N+1 480

VAC FLEX DG (once placed into service), and can feed the steam generators once the plant

is cooled down to below 400 ºF, drawing feedwater suction from the condensate storage pool,

wet cooling tower basins, or refueling water storage pool, and discharging into one of two new

FLEX connections in the emergency feedwater lines. Additionally, the FCCP provides an

additional backup to the charging pumps (one of three which are repowered from the FLEX

DG) for reactor coolant system makeup and cooling.

Further, as described in the licensees FIP Sections 2.3.5.4 and 2.3.10, for an ELAP/LUHS

event that occurs when the plant is in Modes 5 or 6 (with the steam generators not available

for heat removal/core cooling), the FCCP serves as the primary method for making up water

to the reactor coolant system that would start to boil off from the residual decay heat in the

core. In this condition, the FCCP would take suction from the refueling water storage pool

and discharge into the reactor coolant system via one of two new FLEX connections in the

high pressure safety injection lines.

During walkdowns of the reactor auxiliary building -35 ft. elevation, the inspection team noted

that the N FCCP was installed on a pedestal which elevated the base of the pump

approximately 18 inches above the floor grade. The licensee stated that the FCCP was

elevated to account for the potential for internal flooding caused by a seismic event rupturing

a large water source that was not seismically robust. This consideration in the FLEX strategy

was required by NEI 12-06, Revision 0, Section 5.3.3, item 2, which states Consideration

should be given to the impacts from large internal flooding sources that are not seismically

robust and do not require ac power (e.g., gravity drainage from lake or cooling basins for

nonsafety-related cooling water systems). However, the inspection team noted that the

electrical cable connection for the FCCP did not have a waterproof cap, nor was the cable

secured such that the cable end would not drop below the base of the pump, potentially

exposing the electrical connection to damaging internal flood waters.

The inspection team reviewed engineering calculation ECM 15-004, Waterford 3 FLEX

Internal Flooding Calculation, Revision 0, and determined that the licensee had considered

several internal water sources, in particular piping and a tank associated with the fire

protection system, that were determined to not be seismically robust and could contribute to

flooding of the reactor auxiliary building -35 ft. elevation. The licensee considered the critical

maximum flood level on the elevation to be 15 inches, and validated that the non-robust

internal flooding sources would not exceed that elevation. However, the team noted that the

licensees calculation did not consider the FCCP electrical cable connections length and

capability to fall into the potential internal flood waters. The team further reviewed the

sections in FSG-005, Initial Assessment and FLEX Equipment Staging, Revision 6,

associated with the staging and deployment of the FCCP under various conditions, and

determined there was no procedural guidance provided to ensure the assumptions made in

ECM 15-004 calculation were maintained by way of administrative controls.

The team determined that the lack of physical or procedural controls to ensure the FCCP

electrical cable connection met the assumptions of the internal flooding calculations could

affect both the pre-staged N FCCP and the backup N+1 FCCP. However, the team

determined that the N+1 FCCP would only be affected in the situation where the pump was

moved from the storage building and staged at the reactor auxiliary building -35 ft. elevation

13

Enclosure 2

to restore the N-capability of the FCCP prior to the seismic event causing the internal

flooding. While the N+1 FCCP is skid mounted, it had similar critical dimensions and an

electrical cable connection similar to the pre-staged N FCCP.

Corrective Action(s): In response to the inspection teams questions, the licensee

documented the concerns in the corrective action program and initiated actions to secure the

FCCP electrical cable connection to the top of the pump casing to preclude the connection

from falling into the potential internal flood waters following a seismic event. The licensee

also made changes to procedure FSG-005 to include the use of cable stands (procured and

staged for use in a FLEX equipment box in the reactor auxiliary building) to ensure the cable

from the FLEX distribution panel connecting to the FCCP is maintained above potential

internal flood waters.

Corrective Action Reference(s): CR-WF3-2017-07709; CR-WF3-2017-07712

Performance Assessment:

Performance Deficiency: That the station did not give adequate consideration to the impacts

from large internal flooding sources that were not seismically robust was a performance

deficiency. Specifically, the electrical connection to provide power from the FLEX DG to the

FLEX core cooling pump (FCCP) on reactor auxiliary building -35 elevation was not protected

from potential impacts of seismically-induced internal flooding from non-seismic fire protection

piping in the area either by design or by procedure, contrary to NEI 12-06, Revision 0,

Section 5.3.3, item 2. The electrical connection for the FCCP was not water resistant or

otherwise protected from flooding. Following a seismically-induced internal flooding event,

the connection could potentially either be damaged or otherwise fail from water exposure,

potentially rendering the pump unable to function.

Screening: The performance deficiency is more than minor because it is associated with the

equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects

its objective to ensure the availability, reliability, and capability of systems that respond to

initiating events to prevent undesirable consequences (i.e., core damage). Specifically,

without design or procedural protections in place to address the performance deficiency, a

seismically-induced internal flood could render either the N-set or N+1 FCCP unable to fulfill

its functions during an ELAP event.

Significance: The team assessed the significance of the finding using IMC 0609, Appendix O,

Significance Determination Process for Mitigating Strategies and Spent Fuel Pool

Instrumentation (Orders EA-12-049 and EA-12-051), dated October 7, 2016. The team

determined that the performance deficiency did not (1) impact the spent fuel pool

instrumentation order (EA-12-051); (2) did not involve unavailable equipment, deficient

procedures, or deficient training that would result in the complete loss of one or more FLEX

functions (where both the N and N+1 equipment were completely lost) with an exposure

period of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and (3) did not involve programmatic deficiencies that

reduced the effectiveness of the Mitigating Strategies. Further the product of the findings

exposure time and the applicable external event initiating event frequency (i.e., in this case a

seismic event) was not greater than 1E-6. Therefore, the finding was of very low safety

significance (Green).

14

Enclosure 2

Cross-cutting Aspect: The finding had a problem identification and resolution cross-cutting

aspect associated with evaluation, in that the licensee failed to completely consider the

impacts of the potential internal flood height on all electrical components for the FCCP [P.2].

Enforcement: Inspectors did not identify a violation of regulatory requirements associated

with this finding.

Observation

The Waterford 3 site is unique in that there are approximately 15 permanent, large chemical

industry facilities within 5 miles of the station along the Mississippi River. As such, the

licensee has design basis analyses documenting the potential hazards from the wide variety

of flammable and toxic materials produced and/or stored at these industrial facilities described

in Updated Final Safety Analysis Report (UFSAR), Section 2.2.3. Additionally, because of

these hazards, the licensee (1) utilizes Technical Specification-required, broad range gas

detection equipment to detect and initiate automatic isolation of the control room ventilation

system (described in UFSAR, Section 6.4), and (2) established comprehensive toxic chemical

contingency (i.e., emergency) procedures to address actions to ensure the safety of station

personnel during such releases from nearby facilities. The equipment and procedures have

been described in the UFSAR and established since early in the operations of Waterford 3,

well before the development of the FLEX strategies in accordance with Order EA-12-049.

The inspection team reviewed the licensees FLEX strategies and planning to determine if the

licensee had considered whether one of the beyond-design-basis external events that could

affect Waterford 3 would also adversely affect the nearby industrial facilities. The team

considered that the external event may consequently cause those facilities to experience

failures in electrical power and safety systems resulting in a release of toxic gas which could

affect Waterford 3 and the stations ability to implement the FLEX strategies. In particular, the

licensees FLEX strategy includes actions to ventilate the control room envelope by opening

up several doors to the outside environment. Normally this area is operated at a positive

pressure to limit in-leakage and with monitoring using broad range gas detection equipment

that can automatically initiate control room envelope isolation. Additionally, the licensees

FLEX deployment strategy requires operations and other staff to work in the outside

environment to stage, operate, and monitor the FLEX equipment during Phases 1 and 2 of an

ELAP/LUHS event.

The Waterford 3 staff indicated that they had not considered those offsite effects because

NEI 12-06 (Section 2, Boundary Conditions, page 8) did not require licensees to consider

independent, concurrent events in establishing FLEX strategies. However, the inspection

team noted that NEI 12-06 also states in Section 3.2.2, Item 11, that Plant procedures/

guidance should consider accessibility requirements at locations where operators will be

required to perform local manual operations. procedures/guidance should identify the

protective clothing or other equipment or actions necessary to protect the operator, as

appropriate.

While the station maintains personal protective equipment (i.e., self-contained breathing

apparatus (SCBA)) which could be employed to protect staff should an offsite toxic gas

release encroach the site, the inspection team determined that the licensee had not evaluated

whether the supply of breathing air on-site (and other necessary personal protective

consumables) would be sufficient for the duration of a beyond-design-basis event until

15

Enclosure 2

additional offsite resources arrived (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or later). Similarly, the licensee had not

evaluated the potential impacts on the FLEX strategy in opening the control room envelope,

nor variations in the FSG-directed actions to account for the potential concurrent toxic gas

impacts on the site.

Licensee Actions: Considering the conflicting requirements in NEI 12-06, the licensees

design bases SSCs, and the specifics of the industrial environment surrounding the station,

the licensee documented the inspection teams questions in the corrective action program

with an action to convene a cross-discipline team of Operations, Engineering, and EP

[emergency preparedness] to develop and assess potential enhancements to implementation

of the FLEX mitigating strategy to address coincident toxic chemical release during a Beyond

Design Basis Event.

Subsequently, the licensee initiated actions to address the observations and questions,

including:

Established/enhanced processes with the local large chemical industrial facilities and St.

Charles Parish Emergency Operations Center (EOC), whereby the licensee will be able

to obtain the status of the industrial facilities, and any information on potential toxic gas

releases, directly from the St. Charles EOC during beyond-design-basis events

Procurement of additional handheld toxic gas detection equipment

Procurement of additional breathing air resources to support field and control room

FLEX activities during potential on-site impacts from a nearby, concurrent toxic gas

release

Conducting additional GOTHIC thermodynamic analyses to support development of

contingency FLEX strategies for situations where the control room and switchgear rooms

may need to be maintained closed or re-closed (rather than ventilated) due to on-site

impacts from a nearby, concurrent toxic gas release

Staging of additional FLEX personal protective equipment (SCBAs, handheld gas

monitoring equipment, etc.) at locations in the plant near FLEX equipment staging areas

Enhancements to the overall FLEX strategies/FSGs and severe weather preparation

procedures, including triggers to implement toxic gas response contingencies during a

beyond-design-basis event

Evaluation and documentation of the enhancements to the FLEX strategies in

accordance with the guidance in NEI 12-06, Section 11.8, Configuration Control

Additional staff training on the revised FLEX contingency strategies and equipment

Corrective Action References: CR-WF3-2017-07689; CR-WF3-2018-00998

Planned Actions: As of the end of the inspection, the licensee had not completed all of the

actions to address the observations/questions, though they were tracked within the corrective

action program, with anticipated completion in 2018. These actions may be subject to further

NRC inspection.

16

Enclosure 2

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

On September 22, 2017, the inspectors presented the on-site inspection results in a

management debrief to Mr. R. Gilmore, Director, Regulatory and Performance Improvement,

and other members of the site staff.

On June 7, 2018, the inspectors presented the complete inspection results in a telephonic

exit meeting to Mr. J. Dinelli, Vice President, and other members of the site staff.

17

Enclosure 2

DOCUMENTS REVIEWED

TI 2515/191 - Inspection of the Implementation of Mitigation Strategies and Spent Fuel Pool

Instrumentation Orders and Emergency Preparedness Communication/Staffing/Multi-Unit Dose

Assessment Plans

Condition Reports (CR-xx-xxxx)

CR-WF3-2017-7572*

CR-WF3-2017-7610*

CR-WF3-2017-7640*

CR-WF3-2017-7653*

CR-WF3-2017-7662*

CR-WF3-2017-7689*

CR-WF3-2017-7694*

CR-WF3-2017-7709*

CR-WF3-2017-7711*

CR-WF3-2017-7712*

CR-WF3-2017-7722*

CR-WF3-2018-0998*

LO-WLO-2017-0058

- Notes CRs written as a result of the inspection

Procedures

(Number)

Title

Revision

or Date

EN-FAP-EP-010

Severe Weather Response

06

FIG-001

FLEX Implementing Guideline - Extended Loss of AC

Power

004

FSG-001

Long Term Inventory Control

003

FSG-002

Alternate EFW Suction Source

002

FSG-003

Alternate Low Pressure Feedwater

002

FSG-004

ELAP DC Bus Load Shed and Management

002

FSG-005

Initial Assessment and FLEX Equipment Staging

006

FSG-005

Initial Assessment and FLEX Equipment Staging

008

FSG-006

Alternate CSP Makeup

001

FSG-007

Loss of Vital Instrumentation or Control Power

001

FSG-008

Alternate RCS Boration

000

FSG-009

Low Decay Heat Temperature Control

000

FSG-010

Safety Injection Tank Isolation

000

FSG-011

Alternate Spent Fuel Pool Makeup and Cooling

001

FSG-012

Alternate Containment Cooling

000

FSG-013

Transition from FLEX Equipment

001

FSG-100

BDBEE/ELAP Emergency Response Organization FSG

000

FSG-101

BDBEE/ELAP Communications FSG

003

OI-004-000

Move FHB Rounds to RCA Watch

029

OI-004-000

Update order of log points

029

OP-901-513

Spent Fuel Pool Cooling Malfunction

021

OP-901-521

Severe Weather and Flooding

324

OP-902-005

Station Blackout Recovery

021

SAMG-Intro

Introduction

002

S-SAMG-01

Loss of Large Areas of the Plant Due to Fire/Explosion

019

18

Enclosure 2

Work Orders

(Number)

Title

Revision

or Date

00462201-01

FS ILI3000, Perform Functional Test (FLEX Standby

PM)

52697490-01

FLEX Standby PM-1 Year Operational Test-

FLEXMDSG0001

52698032-01

FLEX Standby PM- 6 Month Functional Test-

FLEXMDSG0001

52703445-01

FLEX Standby PM-1 Year Fluid Analysis-

FLEXMDSG0001

52706314-01

FLEX Standby PM- 1 Year Filter Replace/Lube -

FLEXDRE001

6/6/2017

52706797-01

FLEX Standby PM - 1 Year Inventory - Aux Bldg

6/21/2017

52730989-01

FLEX Standby PM- 6 Mo Functional - FLEXMPMP0006

Pump

52730990-01

FLEX Standby PM- 6 Month Functional -

FLEXMPMP0003 Pump

52730991-01

FLEX Standby PM-6 Month Inspection Debris Removal -

FLEXDRE001

6/6/2017

52754363-01

FLEX - Sullivan Air Compressor Elec PM, Check Battery

7/31/2017

52754364-01

FLEX Sullivan Air Compressor Mech PM, Check Fluids

and Run

7/22/2017

52758030-01

FLEX Standby PM- 3 Mth Walkdown/PMP Rotation -

FLEXMPMP0001

52758033-01

FLEX Standby PM- 3 Mth Walkdown/PMP Rotation -

FLEXMPMP0005

52758034-01

FLEX Standby PM- 3 Month Walkdown -

FLEXMPMP0003 Pump

52758035-01

FLEX Standby PM- 3 Month Walkdown -

FLEXMPMP0006 Pump

52767007-01

FLEX Standby PM-1 Month Walkdown Inspection

FLEXDRE001

8/2/2017

52772909-01

FLEX Standby PM-1 Month Walkdown Inspection

FLEXMDSG0001

8/29/2017

Miscellaneous

Documents (Number)

Title

Revision

or Date

95-0019-000

Godwin Installation, Operation and Maintenance

Manual

12

D-G971.0015

FLEX Diesel Generator Vendor Manual

0

EC 77588

N+1 FLEX Diesel Generator Overturning and Sliding

Due to Wind

5/3/2018

Form Number

02-DSY-PM-00111

Blancett B2800 Flow Monitor Programming and

Installation Manual

SD400

Industrial Diesel Generator Set

TD-G200.0085

Goulds Pump Model 3316 Installation, Operation, &

Maintenance Instructions

1

TD-M924.0015

EFP-IL Signal Processor Operators Manual

0

TD-M924.0025

EFP-IL Signal Processor Technical Manual

0

TD-M924.0035

SFP-1 Level Probe Assembly Technical Manual

0

19

Enclosure 2

Miscellaneous

Documents (Number)

Title

Revision

or Date

TD-W120.3085

Westinghouse Molded Case Circuit Breakers Series C,

F-Frame, For Type EHD, FDB, FD, HFD, FDC, DW,

HFW, FWC

2

TRM Section

3/4.13.2

Diverse and Flexible Coping Strategies (FLEX)

Equipment

136

TRM Section

3/4.13.3

FLEX Fluid and Electrical Connections

136

WF3-CS-16-00003

Design Requirments and Vendor Documentation for

FLEX N+1 Storage Building

0

WF3-SA-14-00002

Waterford 3 FLEX Strategy Development

02

Work Standard-FLEX

FLEX Equipment # FLEXMPMP0003 Diesel Driven

Water Transfer Pump Operations

0

Attachment

IMC 0609, Appendix M - Quantitative Evaluation

Waterford 3 N+1 FLEX Diesel Generator Issue

Summary:

The analyst performed a bounding analysis of the subject finding in accordance with Inspection

Manual Chapter (IMC) 0609, Appendix M, Significance Determination Process using

Qualitative Criteria. The analyst quantified the following items that were common to all

initiators:

1. The conditional core damage probability for an unrecoverable loss of offsite power given

that the N FLEX diesel generator was out of service and the N+1 FLEX diesel generator

failed from an external event;

2. The estimated baseline conditional core damage frequency given the N+1 FLEX diesel

generator was successful; and

3. The estimated probability that the N FLEX diesel generator would be out of service at

the time of an external event occurring.

For each external event, the analyst then preformed the following:

1. Estimated the initiating event frequency that would result in an unrecoverable loss of

offsite power;

2. Calculated the baseline incremental conditional core damage probability (ICCDP);

3. Calculated the failure case ICCDP; and

4. Subtracted the baseline from the failure case to determine the combined ICCDP over the

1-year exposure period.

The analyst combined the ICCDPs for each external event initiator to get the total ICCDP for the

performance deficiency. The result was 1.32 x 10-5. This bounds the significance of the finding

to less than Red.

Performance Deficiency:

The inspection team identified that the licensee established their FLEX strategies for the

protection and utilization of the N+1 FLEX Diesel Generator (DG) which were inconsistent with

the requirements of NEI 12-06, Revision 0, and the approved alternatives to the guidance

described in Section 12.5 of the licensees Final Integrated Plan, and Section 3.14 of the NRC

Safety Evaluation, which was a performance deficiency. Specifically, the licensee established

the FLEX strategies such that the N+1 FLEX DG could not be used under all hazards, and

implemented a revised allowed out of service time inconsistent with the NEI 12-06, Revision 0,

and approved alternative guidance.

More Than Minor Determination:

The performance deficiency is more than minor because it is associated with the equipment

performance attribute of the Mitigating Systems Cornerstone and it adversely affected the

associated cornerstone objective to ensure the availability, reliability, and capability of systems

that respond to initiating events to prevent undesirable consequences (i.e., core

damage). Specifically, in establishing the FLEX electrical strategy the licensee failed to ensure

the N capability of the FLEX DGs remained available under all hazards.

2

Attachment

Evaluation using IMC 0609, Appendix O:

Using IMC 0609, Appendix O, Significance Determination Process for Mitigating Strategies and

Spent Fuel Instrumentation, dated 10/07/16, the inspectors performed a screening of the

finding in accordance with Section 0609O-04, Guidance. Based on this screening,

Appendix O directed that the inspection finding be assessed using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process using Qualitative Criteria.

Evaluation using IMC 0609, Appendix M:

In accordance with Section 4.1, Initial Bounding Evaluation, the senior reactor analyst used the

best available quantitative methods and information to bound the risk significance of the finding.

The external initiating events evaluated were those beyond the design basis of the facility as

prescribed by Order EA-12-049 as follows:

Seismic events

External flooding

Storms such as hurricanes, high winds and tornadoes

Extreme snow, ice and cold, and

Extreme heat

The following documents the assumptions used and the bounding evaluation for each of the

external initiators:

Assumptions:

1. The Waterford 3 SPAR Model, Version 8.54 is the best tool for quantifying the

conditional core damage probability related to this degraded condition, including the

unavailability of the N+1 FLEX diesel generator whenever the N FLEX diesel is out of

service.

2. The seismic evaluation procedures delineated in the Risk Assessment of Operational

Events Handbook, Volume 2, External Events, Revision 1.01 are the best available

method of determining the frequency of a seismically-induced loss of offsite power at the

Waterford 3 site.

3. The seismic hazard vector provided in Table 4A-1, Seismic Hazard Vectors for

the 72 SPAR Plants, of the Risk Assessment of Operational Events Handbook,

Volume 2, External Events, Revision 1.01, is the best available information related to

the seismic hazard at the Waterford 3 site.

4. A seismic event at the Waterford 3 site that is large enough to cause a loss of offsite

power by exceeding the fragility of switchyard insulator stacks would result in a loss of

offsite power that is unrecoverable within a 24-hour period.

5. The risk from the failure to have a strategy that protects the N+1 FLEX diesel generator

from all hazards when staged outside the Nuclear Island east watertight doors

impacted the plant from May 31, 2016 to September 21, 2017. Therefore, an exposure

period of 1-year was used as described in the significance determination process.

3

Attachment

6. The probability of the N FLEX diesel generator being out of service for testing or

maintenance can be approximated by taking the total outage time and dividing it by the

total time that the FLEX strategies were in service since the date of full implementation

on May 31, 2016.

7. The maximum precipitation event postulated at the Waterford 3 site would not cause

flooding that would impact the staging of the N+1 FLEX diesel generator unless it was

accompanied by additional precipitation from a major hurricane.

8. The frequency of a postulated failure of the Mississippi river levee at the Waterford 3 site

can be approximated by the failure frequency of an earthen dam less than 50 feet high

that is more than 5 years since closure.

9. A failure of the Mississippi river levee at the Waterford 3 site would result in a loss of

offsite power that is unrecoverable within a 24-hour period.

10. The mean failure frequency of all earthen dams provided in Table 5A-1, Dam Failure

Rates, of the Risk Assessment of Operational Events Handbook, Volume 2, External

Events, Revision 1.01, is the best available information related to a Mississippi river

levee failure at the Waterford 3 site.

11. The estimated frequency of an F2 or larger tornado strike at the Waterford 3 site,

provided by the Office of Nuclear Reactor Research is the best available information

related to strong winds greater than 125 mph at the Waterford 3 site.

12. Any F2 or larger tornado striking the Waterford 3 site would result in a loss of offsite

power that is unrecoverable within a 24-hour period and the failure of the N+1 FLEX

diesel generator when staged outside the nuclear island.

13. The Waterford 3 reactor plant has been operating for approximately 32.8 calendar years.

14. Given that one hurricane-induced loss of offsite power occurred at the Waterford 3 site

during the operating life of the plant, the mean frequency of a hurricane-induced loss of

offsite power can be quantified as 4.6 x 10-2/year.

15. A hurricane-induced loss of offsite power at Waterford 3 would be unrecoverable within a

24-hour period.

16. If the Generac SD400 vendor manual provides a low operating temperature limit for a

subsystem of the generator skid, then the entire diesel generator is capable of operating

at ambient temperatures above this limit.

17. If the Generac SD400 vendor manual provides a maximum operating temperature limit

for a subsystem of the generator skid, then the entire diesel generator is capable of

operating at ambient temperatures below this limit.

18. The optional manufacturers enclosure for the Generac SD400 is installed over the

N+1 FLEX diesel generator skid and is sufficient to prevent the diesel generator from

being disabled by accumulated snow and ice at the Waterford 3 site.

4

Attachment

External Events Conditional Core Damage Probability:

The analyst determined that the only external initiators of concern are those that would

result in an unrecoverable loss of offsite power. Therefore, using the plant-specific

Standardized Plant Analysis Risk Model, Version 8.54, the analyst quantified the

conditional core damage probability (CCDP) for a non-recoverable, weather-related loss

of offsite power. The CCDP was 3.60 x 10-3. Because the SPAR does not model the

FLEX equipment, this CCDP (CCDPcase) was used as the case value for failure of the

FLEX functions. The dominant core damage sequences included a loss of offsite power

with failure of:

1. The Emergency Power Supply System

Operators to Isolate Controlled Bleedoff

Operators to Maintain Reactor Coolant System Subcooling

Operators to Restore Offsite Power in 2 Hours

Operators to Recover Emergency Diesel Generators in 2 Hours

Operators to Manually Control Emergency Feedwater Flow

Operators to Depressurize the Steam Generators

2. The Emergency Power Supply System

The Turbine-Driven Emergency Feedwater Pump

Operators to Restore Offsite Power in 1 Hours

Operators to Recover Emergency Diesel Generators in 1 Hours

3. The Emergency Feedwater System

Baseline Conditional Core Damage Probability:

Given the case CCDP of 3.60 x 10-3, the analyst calculated a baseline CCDP. To

establish a baseline CCDP, the analyst used a screening value of 0.1 consistent with

PRA analyst practices, for the failure of the FLEX equipment, and applied this to the

dominant sequences. The resulting estimated baseline CCDP (CCDPbase)

was 3.60 x 10-4.

Probability of N FLEX Diesel Generator Being Out of Service for Testing or Maintenance:

In accordance with Assumption 6, the analyst calculated the probability that the N FLEX

diesel generator would be out of service. The analyst used the following data:

1. The N FLEX diesel generator was unavailable for a 17-day period from

January 23, 2017 through February 9, 2017, when an alarm annunciated and the

licensee incorrectly interpreted the alarm as an indication problem.

2. The N FLEX diesel generator was unavailable for a 25-day period from

February 9, 2017, through March 6, 2017, because of a failed component.

3. The analyst estimated that quarterly testing of the N FLEX diesel generator

would remove the diesel from service for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during each test.

4. The analyst estimated that annual testing of the N FLEX diesel generator would

remove the diesel from service for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

5. The total time that the FLEX strategy required the N FLEX diesel generator was

from May 31, 2016 when the licensee declared full FLEX implementation, until

the date the inspectors identified the issue on September 21, 2017.

5

Attachment

Based on Item 5, the N FLEX diesel generator was installed for 478 days. The sum of

the outage times for the N FLEX diesel generator was 42.4 days. Therefore, the

probability of the N FLEX diesel generator being out of service (P(Noos)) was calculated

to be 8.87 x 10-2.

The analyst noted that the unavailability of the N FLEX diesel generator was based on

limited information. Therefore, a sensitivity was performed to assess the range of

potential results. The analyst quantified the following 2 cases:

1. Assuming that the 17-day period from January 23, 2017, through February 9,

2017, was an anomaly, the analyst quantified the risk upon removing this period

from the total outage time of the N FLEX diesel generator. The resulting

unavailability was 5.31 x 10-2 which led to an ICCDP of 7.90 x 10-6.

2. Assuming that the performance deficiency could have reasonably increased the

calculated unavailability of the N FLEX diesel generator, the analyst added a

17-day period to the total outage time and quantified the change in risk. The

resulting unavailability was 1.24 x 10-1 which led to an ICCDP of 1.85 x 10-5.

The change in this value results in approximately 40% change in either direction.

Therefore, the sensitivity of the total ICCDP to this assumption is negligible.

Exposure Period:

In accordance with Assumption 5, the exposure period (EXP) during which the plant was

potentially impacted by the performance deficiency was 1 year.

Seismic Events:

The analyst followed the seismic procedures delineated in the Risk Assessment of

Operational Events Handbook, Volume 2, External Events, Revision 1.01. Section 4.0,

Seismic Event Modeling and Seismic Risk Quantification, provides a method of

calculating the frequency of a seismically-induced loss of offsite power (seismic-LOOP)

given the seismic hazard vector for the site. The analyst calculated a frequency of

2.27 x 10-5/year over the entire seismic hazard spectrum.

Using the values quantified above, the analyst calculated a bounding ICCDP for seismic

initiators as follows:

= ICCDPcase - ICCDPbase

Where:

ICCDPcase

= seismic-LOOP * CCDPcase * P(Noos) * EXP

= 2.27 x 10-5/year * 3.60 x 10-3 * 8.87 x 10-2 * 1 year

= 7.26 x 10-9

6

Attachment

And:

ICCDPbase

= seismic-LOOP * CCDPbase * P(Noos) * EXP

= 2.27 x 10-5/year * 3.60 x 10-4 * 8.87 x 10-2 * 1 year

= 7.26 x 10-10

Therefore:

= 7.26 x 10-9 - 7.26 x 10-10

= 6.53 x 10-9

External Flooding:

As stated in the Final Safety Analysis Report, Section 2.4.3.7, the only conditions

conducive to flooding of the Waterford 3 site are heavy precipitation and failure of the

Mississippi river levee. The inspectors determined that a maximum precipitation event

that was not exacerbated by a hurricane would not cause flooding to a height that would

have impacted the N+1 FLEX diesel generator. Therefore, the only external flooding

scenario of impact would be a postulated failure of the Mississippi river levee.

Using best available information, the analyst assumed that a postulated failure of the

Mississippi river levee would have the same frequency as an earthen dam less than

50 feet high and greater than 5 years since construction. According to the Risk

Assessment of Operational Events Handbook, Volume 2, External Events,

Revision 1.01, Table 5A-1, Dam Failure Rates, the mean failure frequency would be

1.634 x 10-4. It is noteworthy that all forms of dams have a failure rate between 1 x 10-4

and 4 x 10-4, even for blue sky events. The analyst assumed that failure of the levee

would result in an unrecoverable loss of offsite power (flood-LOOP) as well as failure of the

N+1 FLEX diesel generator.

Using the values quantified above, the analyst calculated a bounding ICCDP for external

flooding as follows:

= ICCDPcase - ICCDPbase

Where:

ICCDPcase

= flood-LOOP * CCDPcase * P(Noos) * EXP

= 1.63 x 10-4/year * 3.60 x 10-3 * 8.87 x 10-2 * 1 year

= 5.21 x 10-8

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Attachment

And:

ICCDPbase

= flood-LOOP * CCDPbase * P(Noos) * EXP

= 1.63 x 10-4/year * 3.60 x 10-4 * 8.87 x 10-2 * 1 year

= 5.21 x 10-9

Therefore:

= 5.21 x 10-8 - 5.21 x 10-9

= 4.69 x 10-8

High Winds - Tornados:

Using the methods described in Review of Methods for Estimation of High Wind and

Tornado Hazard Frequencies, prepared for the U. S. NRC by the Center for Nuclear

Waste Regulatory Analyses and the Southwest Research Institute, the Office of Nuclear

Reactor Research determined the frequency of exceedance for tornadic winds in excess

of 111 mph (F2 on Fujita Tornado Damage Scale) striking reactor sites in the United

States. The frequency for this storm was 2.41 x 10-5/year for Waterford 3. According to

the Risk Assessment of Operational Events Handbook, Volume 2, External Events,

Revision 1.01, Section 5.2, Scenario Definition and Quantification, weather-related loss

of offsite power initiating events occur with strong winds greater than 125 mph which is

the middle of the F2 scale. As stated in Assumptions 10 and 11, any F2 tornado striking

the Waterford 3 site would result in a loss of offsite power and damage the N+1 FLEX

diesel generator. Therefore, the tornado-induced loss of offsite power frequency

(tornado-LOOP) was estimated to be 2.41 x 10-5/year.

Using the values quantified above, the analyst calculated a bounding ICCDP for high

winds/tornados as follows:

= ICCDPcase - ICCDPbase

Where:

ICCDPcase

= tornado-LOOP * CCDPcase * P(Noos) * EXP

= 2.41 x 10-5/year * 3.60 x 10-3 * 8.87 x 10-2 * 1 year

= 7.69 x 10-9

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Attachment

And:

ICCDPbase

= tornado-LOOP * CCDPbase * P(Noos) * EXP

= 2.41 x 10-5/year * 3.60 x 10-4 * 8.87 x 10-2 * 1 year

= 7.69 x 10-10

Therefore:

= 7.69 x 10-9 - 7.69 x 10-10

= 4.69 x 10-8

High Winds - Hurricanes:

On August 30, 2005, Waterford 3 divorced from offsite power based on instability in the

regional electrical grid as a result of Hurricane Katrina. No other hurricane-induced loss

of offsite power events have occurred during the life of the plant. The analyst noted that

Waterford 3 has operated approximately 32.8 years. Using a Jeffreys non-informative

prior, the analyst estimated the mean frequency of a hurricane-induced loss of offsite

power (hurricane-LOOP) to be 4.6 x 10-2/year. The analyst also reviewed information

provided by the National Oceanic and Atmospheric Administration (NOAA). This

Administration estimated that the return period for a major hurricane striking within

50 miles of the Waterford 3 site was 20 years. This is equivalent to a frequency of

5.0 x 10-2/year. Assuming that a Category 3 or higher hurricane is likely to cause a loss

of offsite power, this NOAA information corroborates the analysts approximation.

Using the values quantified above, the analyst calculated a bounding ICCDP for

hurricane initiators as follows:

= ICCDPcase - ICCDPbase

Where:

ICCDPcase

= hurricane-LOOP * CCDPcase * P(Noos) * EXP

= 4.60 x 10-2/year * 3.60 x 10-3 * 8.87 x 10-2 * 1 year

= 1.46 x 10-5

And:

ICCDPbase

= hurricane-LOOP * CCDPbase * P(Noos) * EXP

= 4.60 x 10-2/year * 3.60 x 10-4 * 8.87 x 10-2 * 1 year

= 1.46 x 10-6

9

Attachment

Therefore: