Text

NRC Confirmatory Action Letter Follow Up Inspection 05000285/2016007
	ML16042A542
Person / Time
Site:	Fort Calhoun
Issue date:	02/11/2016
From:	Sowa J; NRC/RGN-IV/DRP/RPB-D
To:	Marik S; Omaha Public Power District
	Hagar R
References
	EA-13-243 IR 2016007
	Download: ML16042A542 (35)
	v • d • e

[Type here] UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E LAMAR BLVD

ARLINGTON, TX 76011-4511

February 11, 2016

Shane M. Marik, Vice President

Omaha Public Power District

Fort Calhoun Station FC-2-4

P.O. Box 550

Fort Calhoun, NE 68023-0550

SUBJECT: FORT CALHOUN STATION - NRC CONFIRMATORY ACTION LETTER

FOLLOW UP INSPECTION 05000285/2016007

Dear Mr. Marik:

On January 15, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed a

Confirmatory Action Letter follow-up team inspection your Fort Calhoun Station (FCS) and on

January 14, 2016, discussed the results of this inspection with Mr. Todd Tierney and other

members of your staff. The inspection team documented the results of this inspection in the

enclosed inspection report.

During this inspection, the NRC examined activities conducted under your license as they relate

to public health and safety with the Commission's rules and regulations and with the conditions

of your license. Within these areas, the inspection consisted of selected examination of

procedures and representative records, observations of activities, and interviews with

personnel.

This inspection focused on assessing activities related to the implementation of the

commitments described in Confirmatory Action Letter (CAL) EA-13-243, issued December 17,

2013 (ML13351A395). CAL EA-13-243 confirmed the Omaha Public Power Districts (OPPDs)

commitments to ensure the improvements realized during the previous extended outage remain

in place, and performance continues to improve at the facility. Specifically, this inspection

reviewed the action items associated with the subject commitments to determine which could be

closed, and this inspection report describes the results of those reviews.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public

Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosure, and your

response (if any) will be available electronically for public inspection in the NRCs Public

S.Marik -2-

Document Room or from the Publicly Available Records (PARS) component of the NRC's

Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible

from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic

Reading Room).

Sincerely,

/RA/

Jeffrey Sowa, Chief (Acting)

Project Branch D

Division of Reactor Projects

Docket: 50-285

License: DPR-40

Enclosure: NRC Inspection Report 05000285/2016007

w/Attachment: Supplemental Information

ML16042A542

SUNSI Review Non-Sensitive Publicly Available Keyword:

By: RCH Sensitive Non-Publicly Available NRC-002

OFFICE DRP/B DNMS DRP/D DRP/D

NAME DDodson BBaca BHagar JSowa

SIGNATURE /RA/E- RA/E- /RA/ /RA/

DATE 2-10-16 2-9-16 2/11/16 2/11/16

Letter to S.Marik from Jeffrey Sowa, dated February 11, 2016

SUBJECT: FORT CALHOUN STATION - NRC CONFIRMATORY ACTION LETTER

FOLLOW UP INSPECTION 05000285/2016007

DISTRIBUTION:

Regional Administrator (Marc.Dapas@nrc.gov)

Deputy Regional Administrator (Kriss.Kennedy@nrc.gov)

DRP Director (Troy.Pruett@nrc.gov)

DRP Deputy Director (Ryan.Lantz@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

DRS Deputy Director (Jeff.Clark@nrc.gov)

Senior Resident Inspector (Max.Schneider@nrc.gov)

Resident Inspector (Brian.Cummings@nrc.gov)

FCS Site Administrative Assistant (Janise.Schwee@nrc.gov)

Branch Chief, DRP/D (Jeffrey.Sowa@nrc.gov)

Senior Project Engineer, DRP (Bob.Hagar@nrc.gov)

Project Engineer, DRP/D (Jan.Tice@nrc.gov)

RIV Public Affairs Officer (Victor.Dricks@nrc.gov)

NRR Project Manager (Fred.Lyon@nrc.gov)

Team Leader, DRS/TSS (Thomas.Hipschman@nrc.gov)

RIV RITS Coordinator (Marisa.Herrera@nrc.gov)

RIV Regional Counsel (Karla.Fuller@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

RIV Congressional Affairs Officer (Angel.Moreno@nrc.gov)

OEWEB Resource (OEWEB.Resource@nrc.gov)

OEWEB Resource (Sue.Bogle@nrc.gov)

Technical Support Assistant (Loretta.Williams@nrc.gov)

RIV/ETA: OEDO (Raj.Iyengar@nrc.gov)

RIV RSLO (Bill.Maier@nrc.gov)

ACES (R4Enforcement.Resource@nrc.gov)

ROPreports.Resource@nrc.gov

ROPassessment.Resource@nrc.gov

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000285

License: DPR-40

Report: 05000285/2016007

Licensee: Omaha Public Power District

Facility: Fort Calhoun Station

Location: 9610 Power Lane

Blair, NE 68008

Dates: January 11 - 15, 2016

Inspectors: B. Hagar, Senior Project Engineer (Lead)

D. Dodson, Senior Resident Inspector, Wolf Creek Station

B. Baca, Health Physicist

Approved By: Jeffrey Sowa, Chief (Acting)

Branch D, Division of Reactor Projects

-1- Enclosure

SUMMARY

IR 05000285/2016007; 01/11/16 - 01/15/16; Fort Calhoun Station; Confirmatory Action Letter

Follow-up Inspection.

The inspection activities described in this report were performed from January 11-15, 2016, by

two inspectors from the NRCs Region IV office, and during the several weeks preceding

January 11, 2016, by a health physicist in the NRCs Region IV office. During this inspection,

the inspectors did not identify a finding.

The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red),

which is determined using Inspection Manual Chapter 0609, Significance Determination

Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310,

Components Within the Cross-Cutting Areas. Violations of NRC requirements are

dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for

overseeing the safe operation of commercial nuclear power reactors is described in NUREG-

1649, Reactor Oversight Process.

-2-

REPORT DETAILS

4. OTHER ACTIVITIES (OA)

4OA4 Confirmatory Action Letter (CAL) Inspection Activities (92702)

The inspection team assessed and verified certain commitments described in the

Confirmatory Action Letter (CAL) issued December 17, 2013 (ML13351A395). That CAL

stated that it would remain in effect until the NRC verifies that Omaha Public Power

District (OPPD) effectively implements the commitments identified below:

1. OPPD commits to implement those actions detailed in the December 2, 2013, letter

titled, "Integrated Report to Support Restart of Fort Calhoun Station and Post-

Restart Commitments for Sustained Improvement (ML13336A785), associated with

the following areas:

Organizational Effectiveness, Safety Culture, and Safety Conscious Work

Environment

Problem Identification and Resolution

Performance Improvement and Learning Programs

Design and Licensing Basis Control and Use

Site Operational Focus

Procedures

Equipment Performance

Programs

Nuclear Oversight

Transition to the Exelon Nuclear Management Model and Integration into the

Exelon Nuclear Fleet

2. OPPD commits to complete the following actions detailed in the Flooding Recovery

Action Plan: 1.2.3.21, 1.2.3.82, and 4.4.3.1 through 4.4.3.3. These actions entail:

Item 1.2.3.21 - Inspect tank and equipment on demineralized water tank for

damage

Item 1.2.3.82 - Perform independent spent fuel storage installation route load

test

Item 4.4.3.1 - Gather flood response lessons learned through condition report

reviews to determine if procedure or strategy changes should be implemented

Item 4.4.3.2 - Review flood design basis and determine if the 2011 flood event

provides additional information that should drive design basis changes

Item 4.4.3.3 - Implement procedure and strategy changes as indicated by the

lessons learned review conducted

-3-

3. OPPD commits to complete actions 4.5.1.14 and 4.5.1.15 (tracked through

4.5.3.06) detailed in the Flooding Recovery Action Plan, Perform HELB [High

Energy Line Break] analysis of Auxiliary Steam in the Auxiliary Building and

Implement resolution of Auxiliary Steam piping in the Auxiliary Building.

4. OPPD commits to:

Evaluate the structural design margin for the containment internal structures,

and reactor cavity and compartments, and resolve any deficiencies in

accordance with its corrective action program (CAP).

Regarding Beam 22A and Beam 22B in the containment internal structures,

resolve any deficiencies in accordance with the CAP.

Regarding the reactor head stand, prior to the next use of the reactor head

stand, OPPD will evaluate the structural design margin for the head stand and

resolve any deficiencies in accordance with the CAP.

The sections below report the status of these commitments. Specifically,

Section 4OA4.1 reports the status of those actions detailed in the December 2,

2013, letter titled, "Integrated Report to Support Restart of Fort Calhoun Station

and Post-Restart Commitments for Sustained Improvement.

Section 4OA4.2 reports the status of the subject actions detailed in the Flooding

Recovery Action Plan.

Section 4OA4.3 reports the status of the subject commitments associated with

Auxiliary Steam piping in the Auxiliary Building.

Section 4OA4.4 reports the status of the subject containment internal structures.

Section 4OA4.5 lists the action items within the CAL that remain open after this

inspection.

.1 Actions detailed in the December 2, 2013, letter titled, "Integrated Report to Support

Restart of Fort Calhoun Station and Post-Restart Commitments for Sustained

Improvement (ML13336A785)

In the subject letter, OPPD characterized the subject actions as Key Drivers for

Achieving and Sustaining Excellence. The subsections below describe each of those

key drivers. Each subsection begins with a subsection number and the key driver title in

bold text, and within each subsection are one or more parts that correspond to the key

driver action items listed in the subject letter. Each subsection part begins with an

underlined header that includes the item number, the title, and, in parentheses, the Plant

Integrated Improvement Matrix (PIIM) Action Item (AI) number that the licensee used in

the subject letter to identify the key driver action items. Also,

4

if the NRC had previously closed the action item, then within the subsection part

is only a statement that identifies the inspection report (IR) in which the NRC had

previously closed the action item; however,

if the NRC had not previously closed the action item, then within the part are

statements that describe (1) the inspection scope, (2) the most-notable

observations that resulted from inspecting the action item, and (3) the

assessment results.

For the action items inspected by the team, the team verified implementation via the

following activities, as applicable:

verifying that the action item descriptions correspond to the action item descriptions

in Enclosure 3 of OPPDs December 2, 2013, letter;

reviewing documents produced or revised by the action item and/or records resulting

from implementation of the action item;

verifying completion of the action item as scheduled;

assessing the licensees effective use of appropriate performance metrics to

demonstrate performance improvement; and

where applicable, performing independent verification of improved performance.

.1.1 Organizational Effectiveness, Safety Culture and Safety Conscious Work

Environment

Item 1.a: Organizational Effectiveness (2013-0014)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.1.

Item 1.b: Station Safety Culture/Safety Conscious Work Environment (2013-0006)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.1.

.1.2. Problem Identification and Resolution

Item 2.a: Corrective Action Program (CAP) Excellence Plan - Problem Identification

(2013-0055)

Closed in IR 05000285/2015008 (ML15071A115), Section 4OA5.b.2.

Item 2.b: CAP Excellence Plan - Root Cause and Apparent Cause Quality (2013-0065)

Closed in IR 05000285/2015008 (ML15071A115), Section 4OA5.b.1.

Item 2.c: CAP Excellence Plan - Corrective Action Closure (2013-0062)

Closed in IR 05000285/2015008 (ML15071A115), Section 4OA5.b.1.

5

.1.3. Performance Improvement and Learning Programs

Item 3.a: Performance Improvement (2013-0015)

Closed in IR 05000285/2015008 (ML15071A115), Section 4OA5.b.2.

Item 3.b: Human Performance (2013-0061)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.3.

.1.4. Design and Licensing Basis Control and Use

Item 4.a: Design and Licensing Basis (2013-0086)

(1) Inspection Scope

As described in IR 05000285/2014009 (ML14318A886), Section 4OA4.4, the NRC

previously inspected these action items with satisfactory results:

AI 2013-05570-010, Strengthen the Engineering Assurance Group to improve the

oversight of engineering products that affect the design or licensing basis.

AI 2013-05570-025, Complete Phase 2 of the key calculation identification and

improvement process. Phase 2 of the process evaluates the critical calculations

defined purpose and methodology, defined acceptance criteria, and

appropriateness of the results and conclusions.

AI 2013-05570-067, Develop and implement an aggregate station performance

indicator to measure the effectiveness of maintenance and use of licensing and

design bases information.

AI 2013-05570-079, Decide the appropriate Design Basis Document (DBD)

model for Fort Calhoun Station.

AI 2013-05570-091, Perform a technical assessment of modifications performed

between January 1, 1989, and January 1, 2007, on a population of the top six

risk significant systems that provides a 95/95 confidence level that no nuclear

safety issues have been introduced into the plant.

In IR 05000285/2014009, the NRC closed AI 2013-05570-049 (Improve the

engineering support personnel training regarding the design and licensing basis)

with the comment that upon final closure, the NRC would review this action item for

adequacy.

During this inspection, besides reviewing the adequacy of AI 2013-05570-049, the

team also reviewed implementation of these action items:

AI 2013-05570-026, Identify and define the current licensing bases and assure

licensing bases documentation remains current, accurate, complete, and

retrievable.

6

AI 2013-05570-052, Deliver the modified training to the engineering support

personnel.

AI 2013-05570-057, Develop performance metrics to trend and trigger action on

the performance of the use, implementation, and identification of design and

licensing bases issues such as, effective and ineffective 50.59 evaluations, and

procedure inadequacies related to design and licensing bases.

AI 2013-05570-076, Identify and define the design bases and assure design

bases documentation remains current, accurate, complete, and retrievable.

AI 2013-05570-092, Complete Phase 3 of the Key Calculation Project. Phase 3

consists of revising any deficient critical calculation or engineering analysis

identified from Phase 2, as needed.

AI 2013-05570-093, Validate the design and licensing basis has been translated

into plant operation by verifying that the operation, surveillance, and maintenance

of the safety-related components do not compromise the design and licensing

basis.

AI 2013-17439-003, Ensure Design Engineering performs at least one

engineering self-assessment on a risk significant system in 2014.

AI 2013-17439-004, Ensure Design Engineering performs at least one

engineering self-assessment on a risk significant system in 2015.

AI 2013-17439-005, Assign condition reports to ensure Design Engineering

continues to perform an engineering self-assessment on risk significant systems

each year.

(2) Observations and Findings

Inspection of AIs 2013-05570-057, 2013-17439-003, and 2013-17439-004

resulted in no notable observation.

In CR 2013-05570-049, the licensee characterized the action item in a way that

was different from how they had characterized it in their December 2, 2013,

letter. Specifically, in their December 2, 2013, letter, they characterized this

action as:

Modify engineering support personnel initial and continuing training

addressing the design and licensing basis record types and retrieval.

However, in CR 2013-05570-049, they characterized it as:

CAPR-3- Modify the Engineering Support Personnel Training (ESPT) initial

and continuing training programs to incorporate CAPR-1 and CAPR-2.

Training shall include items 1, 2 and 3 from CAPRs 1 and 2 to address the

identification of design and licensing bases, record types that are included,

and the method of retrieval.

7

(In the text above, CAPR stands for corrective action to prevent recurrence.)

Thus, determining the adequacy of CR 2013-05570-049 was beyond the scope

of this inspection, so the NRC deferred that determination until a later inspection.

By completing AIs 2013-17439-003 & -004, the licensee demonstrated that they

can effectively schedule and complete self-assessments of risk significant

systems. This provides confidence that they will complete AI 2013-17439-005 on

its due date.

(3) Assessment Results

The NRC considers AIs 2013-05570-057, 2013-17439-003, 2013-17439-004,

and 2013-17439-005 to be closed.

The NRC deferred inspecting the adequacy of CR 2013-05570-049 until a later

inspection.

At the time of this inspection, the licensee had scheduled completion of the

remaining action items on July 20, 2018:

AI Number Description

2013-05570-026 Identify and define the current licensing bases and assure

licensing bases documentation remains current, accurate,

complete, and retrievable.

2013-05570-076 Identify and define the design bases and assure design

bases documentation remains current, accurate, complete,

and retrievable.

2013-05570-092 Complete Phase 3 of the Key Calculation Project. Phase 3

consists of revising any deficient critical calculation or

engineering analysis identified from Phase 2, as needed.

2013-05570-093 Validate the design and licensing basis has been translated

into plant operation by verifying that the operation,

surveillance, and maintenance of the safety-related

components do not compromise the design and licensing

basis.

Because inspecting the adequacy of CR 2013-05570-049 is not complete and the

activities listed in the table above are not complete, this item remains open.

.1.5. Site Operational Focus

Item 5.a: Site Operational Focus, Operational Decision Making and Anticipating System

Response (2013-0037)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.5.

8

.1.6. Procedures

Item 6.a: Procedure Quality and Procedure Management (2013-0012)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.6.a.

Item 6.b: Abnormal and Emergency Operating Procedures (2013-0031)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.6.b.

Item 6.c: Transition to the Exelon Nuclear Management Model and Integration into the

Exelon Nuclear Fleet (2013-0077)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.6.c.

.1.7. Equipment Performance

Item 7.a: Tornado Protection (2013-0041)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.7.a.

Item 7.b: Equipment Service Life (2013-0088)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.7.b.

Item 7.c: Equipment Reliability/Containment Internal Structures (2013-0013)

(1) Inspection Scope

The team reviewed the implementation of AI 2012-04392-014, Restore the design

criteria for the Internal Structure of Containment, including any needed plant

modifications to beam 22A and B.

(2) Observations and Findings

In CR 2012-04392-014 AI, the licensee characterized this action item as,

Resolve discrepancies for the Internal Structure of Containment, including any

needed plant modifications. Implement design modifications to restore the

Containment Internal Structure (CIS) to within its design basis requirements.

On December 23, 2015, the licensee submitted LIC-15-0142, Supplement of

License Amendment Request 15-03; Revise Current Licensing Basis to Use ACI

Ultimate Strength Requirements, (ML15363A042), to supersede License

Amendment Request (LAR) 15-03, Revise Current Licensing Basis to Use ACI

Ultimate Strength Requirements, dated August 31, 2015. To allow the NRC time to

respond to that request, the licensee extended the due date for CR 2012-04392-014

AI to December 15, 2016.

The licensee is tracking the implementation part of this action item in CR 2012-

04392-045 AI, which they described as,

9

Implement design modifications to restore the Reactor Cavity and

Compartments (RC&C) to within its design basis requirements during the next

refueling outage (RFO 27).

Because resolution of CR 2012-04392-045 AI is linked to the resolution of CR 2012-

04392-014 AI, the licensee also extended its due date to December 15, 2016.

(3) Assessment Results

Because CR 2012-04392-014 AI is not complete, CAL item 7.c remains open.

Item 7.d: Equipment Reliability/Equipment Performance (2013-0027)

(4) Inspection Scope

The team reviewed the implementation of the following action items:

AI 2012-08134-039, Perform an interim effectiveness review of the Plant Health

Committee process and performance.

AI 2012-08134-040, Perform a final effectiveness review of the Plant Health

Committee process and performance.

(5) Observations and Findings

For the interim effectiveness review (EFR) described in AI 2012-08134-039:

One acceptance criterion was to identify no self-assessment area for

improvement related to performance monitoring being consistently

performed.

On December 2, 2014, as documented in Focused Area Self-Assessment

(FASA) RA 2013-1147-004, the licensee declared that the subject EFR had

failed, because although the site had transitioned to Exelon procedure ER

AA2001, Plant Health Committee, insufficient time had passed to fully

implement that procedure and align plant-health-committee behaviors.

Consequently, the licensee scheduled a follow-up interim effectiveness

review via AI 2012-08134-085.

On April 8, 2015, the licensee closed AI 2012-08134-085 with the note that

they had completed the EFR using procedure PI-AA-125-1004,

Effectiveness Review Manual, and had determined that the EFR had been

effective.

For the EFR described in AI 2012-08134-040:

One acceptance criterion was for the Equipment Reliability Index (ERI) to be

in the second quartile of industry performers or better.

On June 25, 2015, the licensee determined that this EFR had failed, because

the sites ERI of 75 was within the industrys fourth quartile. To address this

10

failure, the licensee initiated CR 2015-08257. Via that CR, the licensee

initiated AI 2012-08134-086 to require an EFR using procedure PI-AA-125-

1004 with a replacement acceptance criterion and a due date of January 17,

2017.

(6) Assessment Results

AI 2012-08134-039 is closed because the licensee has completed the action.

AI 2012-08134-040 is closed because:

the actions described in AI 2012-08134-086 are sufficient to fully address the

actions originally described in AI 2012-08134-040,

AI 2012-08134-086 is on schedule such that it will be completed on or before

the due date,

AI 2012-08134-086 involves only a final effectiveness review, and

the licensee has demonstrated that they can successfully complete

effectiveness reviews.

Because both AI 2012-08134-039 and AI 2012-08134-040 are closed, CAL item 7.d

is closed.

Item 7.e: Electrical Equipment Qualification (EEQ)/High Energy Line Break (2013-0021)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.7.e.

Item 7.f: Safety System Functional Failures (2013-0056)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.7.f.

Item 7.g: Cables and Connections (2013-0033)

(1) Inspection Scope

As described in IR 05000285/2014009 (ML14318A886), Section 4OA4.7, the NRC

previously inspected this PIIM Action Item but did not close it. Instead, that

inspection produced several comments that are summarized below. Therefore, the

team reviewed the implementation of the following action items:

AI 2012-08617-011, Provide procedural expectations and guidance to electrical

craft for handling aged electrical cables.

AI 2012-03544-014, Develop a change management plan to implement the

cables and connections program.

AI 2012-08134-026, Execute plans to recover the EEQ and cable aging

management programs.

11

AI 2009-4216-020, Perform an effectiveness review of the strategy for

maintaining dry those safety-related and important-to-safety cables susceptible to

wetting.

AI 2013-17441-001, Complete an assessment report on Cables and Connections

Program.

AI 2013-17441-002, Complete an assessment report on Verification of Material

Condition of Medium & Low Voltage Safety Related Cables Submerged.

(2) Observations and Findings

Inspection of AIs 2012-03544-014, 2013-17441-001, and 2013-17441-002

resulted in no notable observation.

In IR 05000285/2014009, Section 4OA4.7.g, the NRC noted that the licensee

had closed AI 2012-08617-011 without completing some of the required actions.

The licensee initiated CR 2014-06939 to address this issue.

The inspectors review of CR 2014-06939 verified that the licensee had

completed the actions they had previously missed.

In IR 05000285/2014009, Section 4OA4.7.g, the NRC noted that the licensee

had closed AI 2012-08134-026 without ensuring that the program owners for the

Electrical Equipment Qualification program and the Cable Ageing Management

program were properly qualified. In response to the teams observation, the

licensee initiated CR 2014-9499 to address this issue.

The inspectors review of CR 2014-9499 verified that for both of the subject

programs, both of the program owners and both of the backup program owners

now are qualified.

In IR 05000285/2014009, Section 4OA4.7.g, the NRC noted that the licensee

had failed to accurately transcribe the action associated with AI 2009-4216-020

from its December 2, 2014, letter to the NRC (ADAMs Accession

Number ML13336A785) into AI 2009-4216-020. Specifically, the inspectors

noted that instead of completing an effectiveness review of the strategy for

maintaining dry those safety-related and important-to-safety cables susceptible to

wetting, the licensee had ensured that the long-term strategies for the subject

cables and for keeping the subject manholes dry were in place. In response to

the teams notation, the licensee initiated CR 2014-09009 to document that no

action item had implemented the subject effectiveness review.

The inspectors review of AI 2014-09009-002 verified that the licensee had

satisfactorily completed an effectiveness review of the subject strategy and that

the subject review had concluded that effective implementation of the corrective

actions for CR 2009-4216 had protected the subject safety related-cables from

submergence.

12

(3) Assessment Results

The team considers CAL item 7.g closed.

.1.8. Programs

Item 8.a: Engineering Rigor (2013-0011)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.a.

Item 8.b: Equipment Safety Classification and Safety Related Equipment Maintenance

(2013-0036)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.b.

Item 8.c: Electrical Bus Modifications and Maintenance (2013-0016)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.c.

Item 8.d: Deficiencies in Design and Implementation of Fundamental Regulatory

Required Processes (2013-0007)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.d.

Item 8.e: Design Change 10 CFR 50.59 Practices (2013-0066)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.e.

Item 8.f: Piping Code and System Classification and Analysis (2013-0071)

(1) Inspection Scope

As described in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.f, the NRC

previously inspected the following action items with satisfactory results:

AI 2012-07724-022, Review all Class I piping modifications since April 8, 1994,

and document the effectiveness of the procedure for ensuring that thermal

fatigue analysis was performed.

AI 2012-07724-025, Review the United States of America Standard

(USAS) B31.7 and ASME III code reconciliation and correct any code

discrepancies.

During this inspection, the team reviewed the implementation of the following action

items:

AI 2012-07724-023, Provide calculations documenting thermal fatigue analysis

on the Class I piping systems for primary plant sampling, reactor coolant gas

vent, reactor coolant, safety injection, and waste disposal in accordance with

United States of America Standards (USAS) B31.7 Draft 1968.

13

(2) Observations and Findings

IR 05000285/2014009 (ML14318A886), Section 4OA4.8.g stated that the licensee

would not be able to complete AI 2012-07724-023 until the NRC completes its review

of Licensee Amendment Request 14-04.

In response to Licensee Amendment Request 14-04, on August 10, 2015, the NRC

issued Amendment No. 283 to the licensees operating license. This amendment

allows the licensee to perform pipe-stress analyses of non-reactor coolant system

safety-related piping in accordance with the American Society of Mechanical

Engineers Boiler and Pressure Vessel Code, Section Ill, 1980 Edition (no Addenda)

as an alternative to the current Code of Record (i.e., USAS 831.7, 1968 (DRAFT)

Edition).

Subsequently, the licensee assigned to AI 2012-07724-023 a due date of June 1,

2016.

(3) Assessment Results

Until the licensee completes AI 2012-07724-023, CAL Item 8.f remains open.

Item 8.g: Vendor Manual and Vendor Information Control Program (2013-0060)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.g.

Item 8.h: Safeguards Information Digital Storage Control (2013-0009)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.h.

Item 8.i: Operability Determination (2013-0107)

(1) Inspection Scope

As described in IR 05000285/2014009 (ML14318A886), Section 4OA4.8.i and in

05000285/2015008 (ML15071A115), NRC inspection teams reviewed the

implementation of the following action items:

AIs 2013-19752-001, -037; -038; -039; and -040; as part of the quarterly training

curriculum review committee agenda, review operability determination

performance indicators from the Engineering Assurance Group and the

Operability Determination Quality Review Board. This will be a repeated action

through 2014.

AI 2013-19752-002, Conduct oral boards of all operators who make immediate

operability determinations or screen condition reports.

AI 2013-19752-005, Develop interim guidance for resolving unclear operability

references. Include relating the use of prompt operability determinations with

CAP, and current procedure direction, and its level of detail.

14

AI 2013-19752-006, Formalize the Operability Determination Quality Review

Board into a Fort Calhoun Station procedure.

AI 2013-19752-007, Develop a method for ensuring that immediate operability

determinations which fail the minimum Operability Determination Quality Review

Board acceptance criterion (<70% unsupported operability determination) are re-

performed by the On-Shift Crew.

AI 2013-09494-036, Institute a change to NOD-QP-31 (or equivalent Exelon

document) which incorporates clear and complete directions for completion of

each applicable step of supporting process forms.

AI 2013-19752-010, Develop specific guidance that directs personnel screening

plant conditions or equipment failures to ensure actions are taken as required by

the technical specifications (What to do when this fails procedure).

AI 2013-19752-011, Screen the population of Fort Calhoun Station surveillances

and relate these to the associated limiting condition for operations they support.

AI 2013-19752-012, Review existing testing criteria, direction, or methodologies

against industry norms.

AI 2013-19752-013, Review material previously contained in Technical Data

Book (TDB) VIII to ensure it resides in other documents that are clearly linked to

the associated technical specification limiting condition for operations.

AI 2013-19752-021, -022, -023, and -024; Conduct a common factors analysis of

immediate operability determinations quarterly with results and actions approved

by the MRC. Action will be on-going through 2014.

AI 2013-19752-025, -026, -027, and -028; Conduct a common factors analysis of

prompt operability determinations quarterly with results and actions approved by

the MRC. Action will be on-going through 2014.

AI 2013-19752-029, -030, -031, and -032; Present to Plant Review Committee

(PRC) licensee event reports, results of operability determination performance

metrics, and common factor analysis no less than semi-annually. Action will be

on-going through 2014.

AI 2013-19752-033, Immediate Operability Determination Engineering Assurance

Group Assessment Performance Indicator of green with no more than one

immediate operability determinations score greater than 2.0 per month (on

average) for the period of June 1 through December 31, 2014.

AI 2013-19752-034, Immediate Operability Determination Engineering Assurance

Group Failure Rate Performance Indicator of green with no more than one

immediate operability determinations failure per month (on average) for the

period of June 1 through December 31, 2014.

15

AI 2013-19752-035, Operability Determination Quality Review Board Operability

Determination Performance Indicator of green with average Immediate

Operability Determination (IOD)/Immediate Functionality Assessment (IFA) score

> 90% per month for a period of June 1 through December 31, 2014.

AI 2013-19752-036, Operability Determination Quality Review Board Operability

Determination Failure Rate Indicator green with < 1 failure per month (on

average) for a period of June 1 through December 31, 2014.

(2) Observations and Findings

Inspection of these action items resulted in no notable observation.

(3) Assessment Results

The team considers CAL item 8.i closed.

.1.9. Nuclear Oversight

Item 9.a: Nuclear Oversight Effectiveness (2013-0010)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.9.a.

.1.10. Transition to the Exelon Nuclear Management Model and Integration into the

Exelon Nuclear Fleet

Item 10.a: Transition to the Exelon Nuclear Management Model and Integration into the

Exelon Nuclear Fleet (2013-0077)

Closed in IR 05000285/2014009 (ML14318A886), Section 4OA4.10.a

.2 Actions detailed in the Flooding Recovery Action Plan

(1) Inspection Scope

As described in IR 05000285/2014004 (ML14317A777), Section 4OA4.5, NRC

inspectors reviewed the implementation of the following action items:

Action Request (AR) 49712-11, Item 4.4.3.1: Gather flood response lessons

learned through CR reviews to determine if procedure or strategy changes

should be implemented.

AR 49712-13, Item 4.4.3.3: Implement procedure and strategy changes as

indicated by the lessons learned review conducted above.

The team reviewed the implementation of the following action items:

AI 2011-8950-025, Item 1.2.3.21: Inspect tank and equipment on DI tank for

damage

16

AI 2012-16067-002, Item 1.2.3.82: [Independent Spent Fuel Storage Installation;

ISFSI] haul route load test

AI 2012-03366-002, Item 4.4.3.2: Review Flood Design Basis and determine if

the 2011 flood event provides additional information that should drive design

basis changes

(2) Observations and Findings

IR 05000285/2014004 (ML14317A777), Section 4OA4.5, states that the NRC had

no notable observation about AR 49712-11 or AR 49712-13.

Concerning AI 2011-8950-025 and AI 2012-03366-002, the team had no notable

observation.

Regarding AI 2012-16067-002:

This action item described a proof test of the haul route with the actual

Transfer Trailer and the tow vehicle used for moving the ISFSI casks and with

a combined weight greater than 110 percent of the Transfer Trailer and tow

vehicles combined weight with a dry cask and irradiated fuel.

On September 26, 2014, and under work request 216509, the licensee

completed a haul route test using a forklift that weighed approximately 37

percent of the weight of the Transfer Trailer and tow vehicle with a dry cask

and irradiated fuel. The licensee noted that the load traveled at a walking

pace with two engineers continuously monitoring the contact of the forklifts

tires with the haul route. The licensee reported that the engineers did not

observe any sign of washout along the haul route, and did not note any haul

route surface reaction or degradation of any kind.

Also, on July 30, 2015, an NRC inspector walked down the haul route and did

not note any degradation of any kind in either the haul route or the areas

immediately adjacent to the haul route.

The Close Comments associated with AI 2012-16067-002 say, in part:

CR 2012-16067-003 will update RE-RR-DFS-003, Loaded DSC/TSC

From Auxiliary Building to ISFSI Operations, to ensure the appropriate

haul route proofing takes place prior to the next FCS Dry Fuel Storage

Campaign. The current estimated scheduled start date for the next FCS

Dry Fuel Storage Campaign is 2018.

The inspectors reviewed CR 2012-16067-003, and verified that the licensee

has updated RE-RR-DFS-003 as described above; the result was Revision

11 to that procedure.

(3) Assessment Results

The team closed AI 2011-8950-025, AI 2012-03366-002, AR 49712-11, and AR

49712-13.

17

Although the licensee has not yet completed a load test as originally described in AI

2012-16067-002 and as currently described in AI 2012-16067-003,

licensee and NRC visual examinations have not identified any evidence of

haul path degradation, and

Revision 11 to procedure RE-RR-DFS-003 indicates that completing the load

test is a scheduled and required element of the licensees next dry fuel

storage campaign.

Therefore, the Actions detailed in the Flooding Recovery Action Plan are closed.

.3 Actions Associated with Auxiliary Steam Piping in the Auxiliary Building

(1) Inspection Scope

The team reviewed implementation of the following action items:

AR 49722-33, 4.5.1.14: Perform [High Energy Line Break] analysis of Auxiliary

Steam piping in the auxiliary building.

AR 61005, 4.5.1.15: Implement resolution of Auxiliary Steam piping in the

auxiliary building.

(2) Observations and Findings

The licensee closed AR 49722-33 to AR 61005, and completed associated action

item 4.5.1.14 via AI 2011-5244-015, which the licensee characterized as,

Perform analysis or calculation and implement required activities which fully

qualifies the [Auxiliary] Steam and Condensate Return lines in the Intake

Structure.

However, their actions addressed more than just the Intake Structure. Specifically,

as described in an attachment to AI 2011-5244-015,

In calculation FC 08353, Environmental Effects from an Auxiliary Steam &

Condensate Return Line Crack in the Intake Structure, the licensee

determined that the plants designers had routed Auxiliary Steam and

Condensate Return lines through the Service Building to the Intake Structure.

The licensee postulated a High Energy Line Crack (HELC) in each of those

lines, and calculated the resulting environmental effects. They concluded

that the environmental limits in the Raw Water Pump Rooms containing the

Raw Water Pumps (AC-10A/B/C/D) and Strainers (AC-12A/B) (the only

components within the Intake Structure to which Electrical Environmental

Qualification (EEQ) applies) are not exceeded.

Analysis of the Intake Structure was not part of the licensees commitment,

but was part of their response to that commitment.

18

In calculation FC 08350, Environmental Effects from Condensate Return

Line Cracks in the Auxiliary Building, the licensee determined that Auxiliary

Steam lines are routed through rooms 69, 81, and 82. Regarding those

rooms:

o The licensee postulated a HELC in Auxiliary Steam piping in room 69,

and determined that the resulting environment would remain mild.

o The licensee did not evaluate room 81 based on previous analyses of

Main Steam Line Breaks (MSLBs) that occur within room 81, because

the MSLBs are bounding for HELCs with respect to temperature,

pressure, and humidity. The licensee documented their analyses of

MSLBs in design analysis FC07889.

o The licensee did not evaluate room 82 because that room does not

contain any credited safe-shutdown equipment, but noted that

according to design analysis FC07889, room 82 would experience a

harsh environment due to a postulated MSLB in Room 81.

o The licensee evaluated dynamic and wetting effects of Auxiliary

Steam HELCs in analysis EA 13-037, discussed further below.

The units Controlled Access Area Ventilation air supply housing VA-17 is

located in room 69, and the units Uncontrolled Access Area Ventilation air

supply housing VA-19 is located in room 81. A HELC within one of those

housings could result in distribution of increased heat load and humidity to

other areas of the plant via the ventilation system. In calculation FC 08462,

the licensee further evaluated the impact of a HELC in AS piping within both

of those air supply housings, and concluded that the effect of the HELC on

airflow within VA-17 and VA-19 would be negligible. The negligible effect of

the HELC on airflow and the control and capacity of heating coils VA-36A(B)

and VA-43A(B) prompted the licensee to conclude that a postulated HELC

within VA-17 and VA-19 would not have a significant effect on either the

Auxiliary Building Controlled Access Area or Uncontrolled Access Area.

In calculation FC 08384, Environmental Effects from Condensate Return

Line Cracks in the Auxiliary Building, the licensee determined that

Condensate Return lines exist in rooms 4, 6, 19, 26, 30, 31, 56, 69, 81, and

82. The licensee did not evaluate rooms 81 and 82 because of the bounding

environment due to a postulated MSLB in Room 81, as discussed above.

The licensee postulated a HELC in Condensate Return piping in rooms 4, 6,

19, 26, 30, 31, 56, and 69 and analyzed the resulting environments with

respect to EEQ limits. The licensee determined that after the postulated

HELC,

o rooms 4, 6, 19, 26, 56, and 69 remain mild environments, and

o rooms 30 and 31 would become harsh environments due to relative

humidity.

19

The licensee determined that the harsh environments in rooms 30 and 31

were acceptable because those rooms contain no credited safe-shutdown

equipment.

The licensee evaluated dynamic and wetting effects of CR cracks in Rooms

4, 6, 19, 26, 30, 31, 56, 69, 81, and 82 in design analysis EA 13-037 (see

below).

In analysis EA 13-037, Auxiliary Steam and Condensate Return High Energy

Line Break Dynamic and Wetting Effects in the Auxiliary Building, the

licensee concluded that no AS or CR HELC in the Auxiliary Building would

produce dynamic wetting effects that would adversely affect the safe

shutdown of the station. However, the licensee identified the need for and

implemented these design changes:

o EC 64326: Because room 19 contains the station air compressors and

safety related Auxiliary Feedwater (AFW) pumps and because EA 13-

037 determined that water may drip on and thus fail the AFW pumps

as a result of a HELC. The licensee re-routed Condensate Return

piping to traverse through the Turbine Building instead of room 19.

The inspectors walked down the previous and current Condensate

Return piping routes.

o EC 62956: To preclude postulating and evaluating a HELC in the

Emergency Diesel Generator (EDG) rooms, the licensee cut and

capped the Auxiliary Steam supply lines to those rooms. The licensee

also installed electric heaters to replace the steam heaters to which

AS had previously supplied steam.

The inspectors walked down the previous Auxiliary Steam steam line

routes, and observed both the subject caps on the steam lines and

the subject electric heaters.

The licensee completed action item 4.5.1.15 Implement resolution of Auxiliary

Steam piping in the auxiliary building, by implementing the engineering changes

discussed above. As noted above, NRC inspectors walked down the implemented

changes.

(3) Assessment Results

The Actions Associated with Auxiliary Steam Piping in the Auxiliary Building are

closed.

.4 Actions Associated with Containment Internal Structures

(1) Inspection Scope

The team reviewed the implementation of the following action items:

20

Evaluate the structural design margin for the containment internal structures,

and reactor cavity and compartments, and resolve any deficiencies in

accordance with its corrective action program (CAP).

Regarding Beam 22A and Beam 22B in the containment internal structures,

resolve any deficiencies in accordance with the CAP.

Regarding the reactor head stand, prior to the next use of the reactor head

stand, OPPD will evaluate the structural design margin for the head stand and

resolve any deficiencies in accordance with the CAP.

(2) Observations and Findings

The licensee addressed these issues in accordance with the CAP as shown below:

Issue Action Item(s)

Evaluate the structural design margin for the CR 2012-04392-014 AI, Item 4

containment internal structures, and reactor & CR 2012-04392-045 AI

cavity and compartments

Regarding Beam 22A and Beam 22B in the CR 2012-04392-048 AI, Item 4

containment internal structures, resolve any

deficiencies

Regarding the reactor head stand, prior to the CR 2012-04392-049 AI, Item 4

next use of the reactor head stand, OPPD will

evaluate the structural design margin for the

head stand and resolve any deficiencies

The discussions below summarize the results of the inspectors reviews of these

action items.

CR 2012-04392-014 AI: The licensee characterized this action item as,

Resolve discrepancies for the Internal Structure of Containment, including any

needed plant modifications. Implement design modifications to restore the

Containment Internal Structure (CIS) to within its design basis requirements.

The licensees analysis of this issue determined that a LAR was warranted to allow

alternate provisions used in the analyses of the internal structures. Accordingly, on

December 23, 2015, the licensee submitted LAR 15-0142 (ML15363A042). To allow

time for the NRC to respond to that LAR, the licensee set the due date for CR 2012-

04392-014 AI to December 15, 2016.

CR 2012-04392-045 AI: The licensee characterized this action item as,

Implement design modifications to restore the Reactor Cavity and

Compartments (RC&C) to within its design basis requirements during the next

refueling outage (RFO 27).

21

The licensee set the date for this AI also to December 15, 2016, to allow time to

resolve LAR 15-03 and its supplement.

CR 2012-04392-048 AI: The licensee characterized this action item as,

Resolve any identified discrepancies concerning Beams 22A and 22B in the

Containment Internal Structure.

As described in OPPD letter LIC-15-0042, the licensee revised this commitment to,

Regarding Beam 22A and Beam 22B in the containment internal structures, prior

to resuming power operation following Fort Calhoun Station Unit 1 Refueling

Outage 28, OPPD will restore full structural design margin as described in the

Fort Calhoun Station licensing basis.

The licensee extended the due date for this AI to December 15, 2016, due to the

complexity of design and the licensees inability to obtain the necessary pre-

engineered components in sufficient time to complete the activities.

CR 2012-04392-049 AI: The licensee prepared and implemented:

Calculation FC 08389, New RVH Support Frame Analysis and Design,

Revision 1, and

Engineering Change EC 58237, Containment Internal Structure RVH Stand

Area,

The inspectors determined that OPPD evaluated the structural design margin for the

reactor head stand by completing and implementing the analyses associated with

FC08389 and EC 58237.

(3) Assessment Results

CRs 2012-04392-014 AI, 2012-04392-045 AI, and 2012-04392-048 AI remain

open, because the associated actions are not complete.

CR 2012-04392-049 AI is closed.

22

.5 Action Items Within the CAL that Remain Open

Sections 4OA4.1 through 4OA4.4 above identified several action items that remain open

after this inspection. To summarize, the table below lists those items. In this table, the

Ref. column refers to the 4OA4 section of this report that discusses each item. The

table is sorted by Due Date.

Ref. AI Number Description Due Date

4OA4.1.4 2013-05570-049 CAPR-3- Modify the Engineering See note 1

(Item 4.a) Support Personnel Training (ESPT)

initial and continuing training programs

to incorporate CAPR-1 and CAPR-2.

Training shall include items 1, 2 and 3

from CAPRs 1 and 2 to address the

identification of design and licensing

bases, record types that are included,

and the method of retrieval

4OA4.1.8 2012-07724-023 Provide calculations documenting June 1,

(Item 8.f) thermal fatigue analysis on the Class I 2016

piping systems for primary plant

sampling, reactor coolant gas vent,

reactor coolant, safety injection, and

waste disposal in accordance with

USAS B31.7 Draft 1968

4OA4.4 2012-04392-014 Evaluate the structural design margin Dec. 15,

(item 4) & for the containment internal structures, 2016

2012-04392-045 and reactor cavity and compartments,

and resolve any deficiencies in

accordance with its corrective action

program (CAP)

4OA4.4 2012-04392-048 Regarding Beam 22A and Beam 22B in Dec. 15,

the containment internal structures, 2016

resolve any deficiencies in accordance

with the CAP

4OA4.1.4 2013-05570-026 Identify and define the current licensing July 20,

bases and assure licensing bases 2018

documentation remains current,

accurate, complete, and retrievable.

4OA4.1.4 2013-05570-076 Identify and define the design bases July 20,

and assure design bases 2018

documentation remains current,

accurate, complete, and retrievable.

4OA4.1.4 2013-05570-092 Complete Phase 3 of the Key July 20,

Calculation Project. Phase 3 consists 2018

of revising any deficient critical

calculation or engineering analysis

identified from Phase 2, as needed.

23

4OA4.1.4 2013-05570-093 Validate the design and licensing basis July 20,

has been translated into plant operation 2018

by verifying that the operation,

surveillance, and maintenance of the

safety-related components do not

compromise the design and licensing

basis.

Note 1: As discussed in section 4OA4.1.4, the NRC deferred determining the adequacy

of this action item until a later inspection.

4OA5 Other Activities

.1 (Closed) Unresolved Item (URI)05000285/2013008-27, Continuous Monitoring

Capability of Post Accident Main Steam Radiation Monitor RM-064

The NRC opened this unresolved item because NRC inspectors had questioned the

capability of post-accident radiation monitor RM-064 to provide representative

measurements due to the system configuration. Specifically, the inspectors suspected

that the system configuration could represent a failure to ensure continuous effluent

monitoring of the main steam lines following a steam-generator-tube-rupture accident

concurrent with a reactor trip and a loss of offsite power.

RM-064 is the licensees post-accident gaseous effluent release monitor corresponding

to NUREG 0707 Section II.F.1.1 High Range Noble Gas Effluent Monitor. That monitor

provides plant operators and emergency planning agencies with information on plant

releases of noble gases during and following an accident.

Under CR 2013-04442, the licensee performed an engineering technical evaluation that

was based on existing radiological analysis Calculation FC06820 used to analyze the

steam generator accident. In this technical evaluation, the licensee removed many of

the conservative assumptions that they had included in Calculation FC06820. Based on

this evaluation and engineering judgment, the licensee determined that the mixing would

be sufficient and the concentration would be adequate to provide a representative

radiation measurement. However, the inspectors raised several questions about this

conclusion, and, in response, the licensee initiated Condition Reports 2013-04442,

2013-05515, 2013-06267, and 2013-10507 to address those questions.

Subsequently, on August 20, 2013, the licensee completed engineering analysis EA 13-

021 to document their determination that RM-064 has a constant mass flow rate, the

sample is uniformly mixed, sufficient pressure exists to drive the sample through the

detector volume, and the radioactive concentrations measured would be conservative at

maximum safety valve flow rate. In addition, the licensee determined the Geiger Mueller

detector would have a conservative factor of 1.8 - 2.0. The licensee determined a

correction factor of 0.825 was necessary to adjust the conservatism and create a more-

representative sample measurement. The licensee revised the RM-064 primary source

and electronic calibration procedures to incorporate the new correction factor.

The inspectors reviewed the associated condition reports, engineering evaluations, and

drawings, and walked down the RM-064 monitoring system. The inspectors compared

24

the licensees assumptions and calculations with industry standards (ANSI N13.1 and

ASTM D1066), licensee technical specifications, USAR, and regulations. The inspectors

reviewed the licensees postulated accident assumptions and offsite dose calculations

with respect to a steam generator tube rupture concurrent with a reactor trip and loss-of-

offsite-power event. The inspector concluded that the post-accident radiation monitor

RM-064 would provide representative high range gaseous effluent measurements during

a design-basis steam-generator-tube-rupture event concurrent with a reactor trip and a

loss-of-offsite-power event.

.2 (Closed) VIO 2013-011-01, Continued Failure to Classify Intake Structure Sluice Gates

as Safety Class 3

The NRC issued this cited violation of 10 CFR 50, Appendix B, Criterion III, Design

Control, because after the NRC had issued the corresponding non-cited violation in

inspection report 05000285/2012002 (ML13045B055), the licensee had remained in

noncompliance until the inspection documented in inspection report 05000285/2013011

(ML13070A399). To address this violation, the licensee initiated CR 2013-05620.

The inspectors reviewed the Apparent-Cause Evaluation documented in CR 2013-05620

and the corresponding corrective actions. Those corrective actions included upgrading

the sluice gate classification to Limited Critical Quality Element (with NRC approval via

Amendment 282), installing safety-related floodwater inlet valves, and updating the

USAR and station procedures accordingly. The inspectors determined that through

these actions, the licensee restored compliance to 10 CFR 50, Appendix B, Criterion III.

This violation is closed.

.3 (Closed) Licensee Event Report (LER) 2012-021-0, HCV-2987, HPSI Alternate Header

Isolation Valve

On February 9, 2013, the licensee submitted this LER after identifying that valve HCV-

2987, High Pressure Safety Injection Alternate Header Isolation, would not have been

able to fulfill its design safety function because of higher-than-acceptable friction in the

valve packing. The licensees associated root-cause-analysis evaluation, documented in

CR-2012-01601-017, stated that in 2008, the licensee had completed a FlowScan

diagnostic evaluation of HCV-2987 that had showed abnormally high stem-packing

friction, but at the time, the licensee did not compare the value to approved calculations

and did not take actions to correct the issue. Consequently, HCV-2987 remained

inoperable from 2008 until July 16, 2013, when the licensee replaced the existing valve

stem packing with a new configuration having a lower packing-friction coefficient. The

subject report also stated that the root cause of the extended inoperability of HCV-2987

was that the licensee had failed to compare FlowScan data with approved calculations

and take corrective actions. One of their corrective actions to prevent recurrence, which

they designated as corrective action to prevent recurrence number one (CAPR-1), was

to write and implement a new procedure characterized as ER-FC- 410-AD-SETPOINT,

Air-Operated Valve Setpoint Control, Revision 0. This procedure established setpoint

control for all Category-I air-operated valves, and required, in part, that the licensee

review all diagnostic test results for compliance with the setpoint criteria for all diagnostic

tests performed.

25

The licensee discovered this condition while reviewing various design calculations as

part of an extent-of-condition review for the condition documented in CR 2011-9945-006

AI, which was that the licensee had discovered non-conservative pressure-regulator

settings for a different valve.

The inspectors considered that the licensees root-cause-analysis evaluation and CAPR-

1 revealed that until the licensee implemented ER-FC- 410-AD-SETPOINT, Revision 0,

the licensee had failed to establish a procedure that required the licensee to review all

diagnostic test results for compliance with the setpoint criteria for all diagnostic tests

performed. The licensees failure to establish that procedure was a performance

deficiency that constituted a violation of Technical Specification 5.8.1. The inspectors

dispositioned that violation as the licensee-identified violation documented in section

4OA7 below.

LER 2012-021-0 is closed.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On January 14, 2016, the inspectors presented the inspection results to Mr. Todd Tierney and

other members of the licensee staff. The licensee acknowledged the results presented. The

licensee confirmed that the inspectors had returned or destroyed any proprietary information

reviewed.

4OA7 Licensee-Identified Violations

Listed below is one violation of very low safety significance (Green) that was identified by the

licensee and is a violation of NRC requirements, which meets the criteria of the NRC

Enforcement Policy for being dispositioned as a non-cited violation.

Technical Specification 5.8.1 requires in part, that procedures shall be established,

implemented, and maintained covering the applicable procedures recommended in

Appendix A of Regulatory Guide 1.33, Revision 2. That appendix states, in part, that

maintenance that can affect the performance of safety-related equipment should be

performed in accordance with written procedures appropriate to the circumstances.

Contrary to the above, maintenance that can affect the performance of safety-related

equipment was not performed in accordance with written procedures appropriate to the

circumstances. Specifically, maintenance that can affect the performance of safety-

related valves was not performed in accordance with a procedure that required the

licensee to review all diagnostic test results for compliance with the setpoint criteria for

all diagnostic tests performed.

As described in CR-2012-01601-017, the licensee restored compliance by writing and

implementing procedure ER-FC- 410-AD-SETPOINT, Air-Operated Valve Setpoint

Control, Revision 0. This procedure requires, in part, that the licensee review all

diagnostic test results for compliance with the setpoint criteria for all diagnostic tests

performed.

26

The licensees failure to complete maintenance that can affect the performance of

safety-related valves in accordance with written procedures appropriate to the

circumstances was a performance deficiency that is more-than-minor because it

adversely affected the Procedure Quality attribute of the Mitigating Cornerstone

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

initiating events to prevent undesirable consequences. Specifically, this performance

deficiency resulted in valve HCV-2987, High Pressure Safety Injection Alternate Header

Isolation, being not able to fulfill its design safety function from February, 2013, through

July, 2013. Using Manual Chapter 0609, Attachment 4, Initial Characterization of

Findings, dated June 19, 2012, the inspectors determined that the finding should be

processed through Manual Chapter 0609, Appendix A, The Significance Determination

Process (SDP) for Findings At-Power, dated July 1, 2012. Using Appendix A, Exhibit 2,

Mitigating Systems Screening Questions, the inspectors determined that the finding

was not a design or qualification deficiency but represented a loss of train function for

greater than the outage time allowed by Technical Specifications. Therefore, a Region

IV senior reactor analyst performed a detailed risk evaluation in accordance with Manual

Chapter 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The analyst

determined that the condition of valve HCV-2987 inoperability would affect only the

plants response to a large-break loss-of-coolant accident followed by the failure of the

instrument air system. The analyst calculated the initiating-event frequency to be 2.63 x

10-10 /year. Also, the analyst determined that the finding did not affect external initiator

risk and would not involve a significant increase in the risk of a large, early release of

radiation. Therefore, this violation has very low (Green) safety significance.

27

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

R. Beck, Manager Chemistry

B. Blome, Manager Site Regulatory Assurance

C. Cameron, Supervisor, Regulatory Compliance

J. Cate ,Manager, Engineering Special Programs

M. Frans, Manager, Special Assignment

C. Gotschall, Corrective Action Program Coordinator

E. Matzke, Senior Regulatory Assurance Engineer

T. Tierney, Plant Manager

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Closed

05000285/2013008-27 URI Continuous Monitoring Capability of Post-Accident Main Steam

Radiation Monitor RM-064

05000285/2013011-01 VIO Continued Failure to Classify Intake Structure Sluice Gates as

Safety Class 3

05000285/2012-021-0 LER HCV-2987, [High Pressure Safety Injection] Alternate Header

Isolation Valve

A1-1 Attachment

LIST OF DOCUMENTS REVIEWED

4OA4 Confirmatory Action Letter (CAL) Inspection Activities

Procedures

Number Title Revision

AOP-01 Acts of Nature 46

CC-AA-201 Plant Barrier Control Program 10

CC-FC-309- Piping Design 0

1011-AD-MEI-8

PED-CSS-3 Procuring, Applying and Inspecting Protective Coatings Inside 7

Reactor Containment Building

RE-RR-DFS-003 Loaded DSC/TSC From Auxiliary Building to ISFSI Operations 11

Condition Reports (CRs)

2012-03366 2012-04392 2012-07724 2014-06939 2014-08532

2014-9499

Action Items (AIs)

2012-04392-014 2012-04392-034 2012-04392-045 2012-04392-048 2012-04392-049

Action Requests (ARs)

49712 61014

Calculations

Number Title Revision

FC08350 Environmental Effects from an Auxiliary Steam Line Crack in 13

Room 69

FC08384 Environmental Effects from Condensate Return Line Cracks in 11

the Auxiliary Building

ML16042A542

Text

Navigation menu

Search