Text

NRC Problem Identification and Resolution Inspection Report 05000458/2017009 and Notice of Violation
	ML17160A401
Person / Time
Site:	River Bend
Issue date:	06/09/2017
From:	Thomas Hipschman; Division of Reactor Safety IV
To:	Maguire W; Entergy Operations
	Hipschman T
References
	IR 2017009
	Download: ML17160A401 (28)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD.

ARLINGTON, TX 76011-4511

June 9, 2017

Mr. William F. Maguire

Site Vice President

Entergy Operations, Inc.

River Bend Station

5485 US Highway 61N

St. Francisville, LA 70775

SUBJECT: RIVER BEND STATION - NRC PROBLEM IDENTIFICATION AND

RESOLUTION INSPECTION REPORT 05000458/2017009 AND NOTICE OF

VIOLATION

Dear Mr. Maguire:

On April 28, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a problem

identification and resolution inspection at your River Bend Station. The NRC inspection team

discussed the results of this inspection with Mr. Marvin Chase, Director, Regulatory &

Performance Improvement, and other members of your staff. The results of this inspection are

documented in Enclosure 2.

The NRC inspection team reviewed the stations corrective action program and the stations

implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating

and correcting problems, and to confirm that the station was complying with NRC regulations

and licensee standards for corrective action programs. Based on the samples reviewed, the

team determined that your staffs performance in each of these areas adequately supported

nuclear safety.

The team also evaluated the stations processes for use of industry and NRC operating

experience information and the effectiveness of the stations audits and self-assessments.

Based on the samples reviewed, the team determined that your staffs performance in each of

these areas adequately supported nuclear safety.

Finally, the team reviewed the stations programs to establish and maintain a safety-conscious

work environment, and interviewed station personnel to evaluate the effectiveness of these

programs. Based on the teams observations and the results of these interviews the team found

no evidence of challenges to your organizations safety-conscious work environment. Your

employees appeared willing to raise nuclear safety concerns through at least one of the several

means available.

The enclosed report discusses a Severity Level IV violation associated with a finding of

very low safety significance (Green). The NRC evaluated this violation in accordance

Section 2.3.2.a of the NRC Enforcement Policy, which can be located at

http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html. The violation is cited in

W. Maguire 2

the enclosed Notice of Violation (Notice) and the circumstances surrounding it are described in

detail in the subject inspection report. The violation is being cited because the licensee

failed to restore compliance with a Title 10 of the Code of Federal Regulations (10 CFR) 50.59

violation associated with the failure to obtain a license amendment that resulted in a more than

minimal increase in the frequency of occurrence of an accident previously evaluated in the final

safety analysis report when implementing a design change to the reactor core isolation cooling

injection location. The NRC previously identified this violation as non-cited violation

(NCV)05000458/2015007-02.

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 is necessary to ensure your compliance with

regulatory requirements.

If you contest the violation or significance of the violation, you should provide a response within

30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear

Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with

copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the

NRC resident inspector at the River Bend Station.

If you disagree with the cross-cutting aspect assignment 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 River Bend Station.

This letter, its enclosure, and your response (if any) 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 Jeffrey Clark for/

Thomas R. Hipschman, Team Leader

Inspection Program and Assessment Team

Division of Reactor Safety

Docket No. 50-458

License No. NPF-47

Enclosure 1: Notice of Violation

Enclosure 2: Inspection Report 05000458/2017009

w/ Attachments: Supplemental Information &

Information Request

cc w/ encl: Electronic Distribution

NOTICE OF VIOLATION

Entergy Operations, Inc. Docket No. 50-458

River Bend Station License No. NPF-47

During an NRC inspection conducted April 10 - 28, 2017, a violation of NRC requirements was

identified. In accordance with the NRC Enforcement Policy, the violation is listed below:

10 CFR 50.59(c)(2) requires, in part, that a licensee shall obtain a license amendment

pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment

if the change, test, or experiment would result in more than a minimal increase in the

frequency of occurrence of an accident previously evaluated in the final safety analysis

report (as updated).

Contrary to the above, as of April 28, 2017, the licensee failed to obtain a license

amendment pursuant to 10 CFR 50.90 prior to implementing a change, test, or

experiment that resulted in a more than minimal increase in the frequency of occurrence

of an accident previously evaluated in the final safety analysis report (as updated).

Specifically, on July 3, 1999, the licensee implemented a design change to the reactor

core isolation cooling injection location from the reactor vessel head to a feedwater line,

but failed to correctly evaluate that a spurious reactor core isolation cooling actuation

into the feedwater line resulted in a more than minimal increase in the frequency of

occurrence of the loss of feedwater heating accident previously evaluated in the updated

final safety analysis report.

This is a Severity Level IV violation (NRC Enforcement Policy Section 6.1.d.2).

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, Texas 76011, and a copy to the NRC

resident inspector at the River Bend Station, within 30 days of the date of the letter transmitting

this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of

Violation, and should include for each 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.

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.

Enclosure 1

Your response will be made available electronically for public inspection in the NRC Public

Document Room or from the NRCs document 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, proprietary, or safeguards 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 9th day of June 2017

2

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Dockets: 05000458

License: NPF-47

Report: 05000458/2017009

Licensee: Entergy Operations, Inc.

Facility: River Bend Station

Location: 5485 U.S. Highway 61N

St. Francisville, LA 70775

Dates: April 10 through April 28, 2017

Team Lead: R. Azua, Senior Reactor Inspector

Inspectors: H. Freeman, Senior Reactor Inspector

P. Jayroe, Project Engineer

B. Parks, Resident Inspector

Approved By: T. Hipschman, Team Leader

Inspection Program and Assessment Team

Division of Reactor Safety

Enclosure 2

SUMMARY

IR 05000458/2017009; 04/10/2017 - 04/28/2017; River Bend Station; Problem Identification

and Resolution (Biennial)

The inspection activities described in this report were performed between April 10 and April 28,

2017, by three inspectors from the NRCs Region IV office and the resident inspector at the

River Bend Station. The report documents one finding of very low safety significance (Green).

This finding involved a violation of NRC requirements; this violation was determined to be

Severity Level IV under the traditional enforcement process. 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, Aspects 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.

Assessment of Problem Identification and Resolution

Based on its inspection sample the team concluded that the licensee maintained a corrective

action program in which individuals generally identified issues at an appropriately low threshold.

Once entered into the corrective action program, the licensee generally evaluated and

addressed these issues appropriately and timely, commensurate with their safety significance.

The licensees corrective actions were generally effective, addressing the causes and extents of

condition of problems.

The licensee appropriately evaluated industry operating experience for relevance to the facility

and entered applicable items in the corrective action program. The licensee incorporated

industry and internal operating experience in its root cause and apparent cause evaluations.

The licensee performed effective and self-critical nuclear oversight audits and self-assessments.

The licensee maintained an effective process to ensure significant findings from these audits

and self-assessments were addressed. However, the team identified a potential weakness in

the stations timeliness for processing certain 10 CFR Part 21 notifications through the operating

experience and corrective action programs. The licensee acknowledged this potential

weakness and indicated their plan to address this through the Entergy fleet.

The licensee maintained a safety-conscious work environment in which personnel were willing

to raise nuclear safety concerns without fear of retaliation.

Cornerstone: Initiating Events

Green. The NRC identified a Severity Level IV violation for the licensees failure to restore

compliance for a non-cited violation (NCV) associated with failure to obtain NRC approval

prior to making a change to the reactor core isolation cooling injection point. Specifically, as

of April 28, 2017, the licensee had not restored compliance with a violation the NRC

identified on October 8, 2015. This violation described a previously made change to the

facility without prior NRC approval in violation of 10 CFR 50.59, Changes, Tests, and

Experiments. The team determined that the licensees failure to restore compliance within

a reasonable amount of time was a performance deficiency. Title 10 CFR 50, Appendix B,

Criterion XVI, requires in part that, measures shall be established to assure that conditions

2

adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material

and equipment, and nonconformances are promptly identified and corrected. The

licensee entered this issue into their corrective action program as Condition Report

CR-RBS-2017-03505.

The finding was more than minor because it is associated with the initiating events aspect of

the reactor safety cornerstone and affected the cornerstone objective to limit the likelihood

of events that upset plant stability and challenge critical safety functions during power

operations. The finding is of very low safety significance (Green) because it did not cause a

reactor trip and the loss of mitigation equipment relied upon to transition the plant from the

onset of the trip to a stable shutdown condition. The finding has a human performance

cross-cutting aspect associated with procedural adherence because individuals failed to

follow the procedures delineated by the corrective action program [H.8]. Originally, the

licensee met the criteria for dispositioning the issue (50.59) as a NCV. However, based

upon the fact that the condition report, which documented the NCV, was closed without

restoring compliance, the licensee no longer met the criteria for a NCV and therefore, this

violation is being cited in a notice of violation (4OA2.5).

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REPORT DETAILS

4. OTHER ACTIVITIES (OA)

4OA2 Problem Identification and Resolution (71152)

The team based the following conclusions on a sample of corrective action documents that were

open during the assessment period, which ranged from July 12, 2015, to the end of the on-site

portion of this inspection on April 27, 2017.

.1 Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The team reviewed approximately 200 condition reports (CRs), including associated root

cause analyses and apparent cause evaluations, from approximately 20,000 that the

licensee had initiated or closed between July 12, 2015, and April 27, 2017. The majority

of these (approximately 20,000) were lower-level condition reports that did not require

cause evaluations. The inspection sample focused on higher-significance condition

reports for which the licensee evaluated and took actions to address the cause of the

condition. In performing its review, the team evaluated whether the licensee had

properly identified, characterized, and entered issues into the corrective action program,

and whether the licensee had appropriately evaluated and resolved the issues in

accordance with established programs, processes, and procedures. The team also

reviewed these programs, processes, and procedures to determine if any issues existed

that may impair their effectiveness.

The team reviewed a sample of performance metrics, system health reports, operability

determinations, self-assessments, trending reports and metrics, and various other

documents related to the licensees corrective action program. The team evaluated the

licensees efforts in determining the scope of problems by reviewing selected logs, work

orders, self-assessment results, audits, system health reports, action plans, and results

from surveillance tests and preventive maintenance tasks. The team reviewed daily

CRs and attended the licensees CR screening meetings and Plant Review Group

(PRG) meetings to assess the reporting threshold and prioritization efforts, and to

observe the corrective action programs interfaces with the operability assessment and

work control processes. The teams review included an evaluation of whether the

licensee considered the full extent of cause and extent of condition for problems, as well

as a review of how the licensee assessed generic implications and previous occurrences

of issues. The team assessed the timeliness and effectiveness of corrective actions,

completed or planned, and looked for additional examples of problems similar to those

the licensee had previously addressed. The team conducted interviews with plant

personnel to identify other processes that may exist, where problems may be identified

and addressed outside the corrective action program.

The team reviewed corrective action documents that addressed past NRC-identified

violations to evaluate whether corrective actions addressed the issues described in the

inspection reports. The team reviewed a sample of corrective actions closed to other

corrective action documents to ensure that the ultimate corrective actions remained

appropriate and timely. The team reviewed a sample of condition reports where the

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licensee had changed the significance level after initial classification to determine

whether the level changes were in accordance with station procedures and that the

conditions were appropriately addressed.

The team considered risk insights from both the NRCs and the River Bend Stations risk

models to focus the sample selection and plant tours on risk-significant systems and

components. The team focused a portion of its sample on the control building heating

and ventilation (HVK) system and automatic depressurization system, which the team

selected for a five-year in-depth review. The team conducted walk-downs of the HVK

system and other plant areas to assess whether licensee personnel identified problems

at a low threshold and entered them into the corrective action program. In addition, the

team also reviewed the licensees use of operational experience and the 10 CFR Part 21

(Part 21) process with respect to these systems.

b. Assessments

1. Effectiveness of Problem Identification

During the 22-month inspection period, licensee staff generated approximately

20,000 condition reports. The team determined that most conditions that required

generation of a condition report per Procedure EN-LI-102, Corrective Action

Program, were entered appropriately into the corrective action program. However,

the Team identified a few errors in the development and processing of CRs:

These errors included assigning the wrong priority to a CR, and/or closing CRs to

a lesser CR, contrary to plant procedures. In most of these instances, the

subsequent actions taken to correct these issues were appropriate to the higher

priority designation. One instance was noted where actions taken were not

commensurate with the required actions for a condition adverse to quality.

Specifically, Condition Report CR-RBS-2015-7259, which was issued in

response to a 10 CFR 50.59 NCV. The licensee closed the CR to a Licensing

Action Request process, which was not an approved process in the corrective

action program. This failure by the licensee to follow their process contributed to

the failure to address the issue in a timely manner, which resulted in a cited

violation. (Section 4OA2.5)

Overall, the team concluded that the licensee generally maintained a low threshold

for the formal identification of problems and entry into the corrective action program

for evaluation. Licensee personnel initiated over 760 CRs per month during the

inspection period. Most of the personnel interviewed by the team understood the

requirements for condition report initiation; most expressed a willingness to enter

newly identified issues into the corrective action program at a very low threshold.

2. Effectiveness of Prioritization and Evaluation of Issues

The sample of CRs reviewed by the team focused primarily on issues screened by

the licensee as having higher-level significance, including those that received cause

evaluations, those classified as significant conditions adverse to quality, and those

that required engineering evaluations. The team also reviewed a number of

condition reports that included or should have included immediate operability

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determinations to assess the quality, timeliness, and prioritization of these

determinations.

Based on the walk-down of the risk-significant systems selected for the five-year in-

depth review, the team observed that the material condition of these systems

appeared to be adequate. With regard to the HVK system, the team noted that the

plant had experienced a number of recurring issues with this system, over several

years, where corrective actions appeared to have been previously ineffective. This

was one of the reasons the team selected this system for review.

The teams focused review of the licensees more recent actions, with regard to the

HVK system, indicated a more rigorous effort was being applied by the licensee to

get this issue under control. One such action was the licensees placement of this

system in their Top Ten Equipment Reliability Action Plan, which focuses more plant

resources to listed systems in an effort to correct identified problems. Interviews with

licensee staff also indicate that a more concerted effort was being made by plant

management to address identified problems with the HVK system with more

permanent solutions. Having said that, this effort is in its nascent stage. The NRC

will continue to monitor these systems to ascertain the effectiveness of the licensees

corrective actions over time.

Overall, the team determined that the licensees process for screening and

prioritizing issues that had been entered into the corrective action program,

supported nuclear safety. The licensees operability determinations were consistent,

accurately documented, and completed in accordance with procedures.

3. Effectiveness of Corrective Actions

Overall, the team concluded that the licensee generally identified effective corrective

actions for the problems evaluated in the corrective action program. The licensee

generally implemented these corrective actions in a timely manner, commensurate

with their safety significance, and reviewed the effectiveness of the corrective actions

appropriately.

The team identified that since early 2016, there has been a reduction in the number

of adverse events caused by human performance errors and work management

deficiencies on the part of River Bend Station employees. However, this positive

data point was tempered by the team having noticed that a potential area of

vulnerability may still exist in the area of supplemental (contract) employees, for

example transmission and distribution personnel. The licensee stated that they were

aware of this vulnerability and indicated were working to address it.

.2 Assessment of the Use of Operating Experience

a. Inspection Scope

The team examined the licensees program for reviewing industry operating experience,

including reviewing the governing procedures. The team reviewed a sample of eight

industry operating experience communications and the associated site evaluations to

assess whether the licensee had appropriately assessed the communications for

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relevance to the facility. The team also reviewed assigned actions to determine whether

they were appropriate.

b. Assessment

Overall, the team determined that the licensee appropriately evaluated industry

operating experience for its relevance to the facility. Operating experience information

was incorporated into plant procedures and processes as appropriate. The licensee was

effective in implementing lessons learned through operating experience. They took full

advantage of being part of the Entergy fleet, to give a thorough review of the operational

experience from a variety of sources. Licensee personnel ensured that significant

issues were dealt with in a thorough and timely manner. This was also true for the

Part 21 process that is within the licensees operational experience program.

The team further determined that the licensee appropriately evaluated industry operating

experience when performing root cause analysis and apparent cause evaluations. The

licensee appropriately incorporated both internal and external operating experience into

lessons learned for training and pre-job briefs.

The team identified one potential weakness with respect to the timeliness of review of

Part 21 notices. Specifically, with regard to Part 21s received where the River Bend

Station was not identified as being affected. In these instances, the notices would be

sent to Entergys Corporate Supply office to be researched. If during this research the

Part 21 was determined to apply to the River Bend Station, the information would be

returned to the site. However, there were no further timeliness goals regarding when

this item should be entered into the corrective action program. The team identified four

examples where the time that lapsed between the publication of a Part 21 notice and

entry into the corrective action program was excessive (80 days to 105 days). None of

the examples identified had an adverse impact on the safety of the plant. Following

discussions with the licensee staff, they acknowledged this insight and indicated that

they plan to address it through the Entergy fleet. The licensee entered this issue into

their corrective action program (Condition Reports CR-RBS-2017-03549 and

CR-HQN-2017-00617).

.3 Assessment of Self-Assessments and Audits

a. Inspection Scope

The team reviewed a sample of licensee self-assessments and audits to assess whether

the licensee was regularly identifying performance trends and effectively addressing

them. The team also reviewed audit reports to assess the effectiveness of assessments

in specific areas. The specific self-assessment documents and audits reviewed are

listed in Attachment 1.

b. Assessment

Overall, the team concluded that the licensee had an effective self-assessment and audit

process. The team determined that self-assessments were self-critical and thorough

enough to identify deficiencies.

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.4 Assessment of Safety-Conscious Work Environment

1. Inspection Scope

The team interviewed 26 individuals in five focus groups. The purpose of these

interviews was: (1) to evaluate the willingness of licensee staff to raise nuclear safety

issues, either by initiating a condition report or by another method, (2) to evaluate the

perceived effectiveness of the corrective action program at resolving identified problems,

and (3) to evaluate the licensees safety-conscious work environment (SCWE). The

focus group participants included personnel from Engineering, Maintenance

(Mechanical, Electrical, and Instrumentation and Controls), Security and Supplemental

(Contract) Engineering personnel. At the teams request, the licensees regulatory

affairs staff selected the participants blindly from these work groups, based partially on

availability. To supplement these focus group discussions, the team interviewed the

Employee Concerns Program manager to assess her perception of the site employees

willingness to raise nuclear safety concerns. The team reviewed the Employee

Concerns Program case log and select case files. The team also reviewed the minutes

from the licensees most recent safety culture monitoring panel meetings.

2. Assessment

1. Willingness to Raise Nuclear Safety Issues

All individuals interviewed indicated that they would raise nuclear safety concerns.

All felt that their management was receptive to nuclear safety concerns and was

willing to address them promptly. All of the interviewees further stated that if they

were not satisfied with the response from their immediate supervisor, they had the

ability to escalate the concern to a higher organizational level. Most expressed

positive experiences after raising issues to their supervisors. All expressed positive

experiences documenting most issues in condition reports.

The team questioned focus group participants whether they were able to submit a

condition report anonymously. Most individuals were aware that they could submit

condition reports anonymously, and were knowledgeable of the process. The team

noted that the number of anonymous CRs has dropped over the last year. This, in

conjunction with the positive staff comments during interviews, was considered an

indicator of improving personnel confidence in the plant and plant management.

2. Employee Concerns Program

All interviewees were aware of the Employee Concerns Program. Most explained

that they had heard about the program through various means, such as posters,

training, presentations, and discussion by supervisors or management at meetings.

All interviewees stated that they would use Employee Concerns if they felt it was

necessary. All expressed confidence that their confidentiality would be maintained if

they brought issues to Employee Concerns.

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4. Preventing or Mitigating Perceptions of Retaliation

When asked if there have been any instances where individuals experienced

retaliation or other negative reaction for raising issues, all individuals interviewed

stated that they had neither experienced nor heard of an instance of retaliation,

harassment, intimidation or discrimination at the site. The team determined that

processes in place to mitigate these issues were being successfully implemented.

Responses from the focus group interviewees indicate that they believe that

management has established and promoted a safety-conscious work environment

where individuals feel free to raise safety concerns without fear of retaliation.

Overall, employees indicated that they have noticed an improved culture on-site. As

described, there was a sense that management is more interested now in addressing

issues in a manner that will result in more lasting solutions. They indicated that there

is more management support for their efforts.

.5 Findings

Failure to restore compliance for a 10 CFR 50.59 Violation

Introduction. The team identified a Green, Severity Level IV, violation for the licensees

failure to restore compliance for a NCV associated with the licensees failure to obtain

NRC approval prior to making a change to the reactor core isolation cooling injection

point. Specifically, as of April 28, 2017, the licensee had not restored compliance with

the NCV the NRC identified on October 8, 2015. This violation described a change,

which was previously made to the facility without prior NRC approval in violation

of 10 CFR 50.59, Changes, Tests, and Experiments, because the evaluation did not

provide adequate justification as to why the change did not result in a more than minimal

increase in the frequency of occurrence of an accident previously evaluated in the final

safety analysis report.

Description. In 1998, the licensee modified the reactor core isolation cooling injection

point from the reactor head spray nozzle to the 'A' feedwater line via the 'A' residual heat

removal shutdown cooling return line. At that time, the licensees evaluation stated that

the modification did not increase the probability of occurrence of an accident evaluated

previously in the Safety Analysis Report (SAR) and, as a result, did not represent an

unreviewed safety question which would have required NRC approval.

In October 2015 the NRC reviewed the licensees modification to the reactor core

isolation cooling injection point as one of the samples during an inspection on

evaluations of changes, tests, and experiments and permanent plant modifications

(Inspection Procedure 71111.17T). The NRC determined that the licensees evaluation

for this modification was inadequate because the licensee had failed to correctly

evaluate that a spurious reactor core isolation cooling actuation injecting through the

feedwater line would also result in the same characteristics, (and therefore increase the

probability of occurrence) of another accident previously evaluated (loss of feedwater

heating) and that this would be more than a minimal increase in frequency.

The requirements governing the authority of production and utilization facility licensees

to make changes to their facilities without prior NRC approval are contained in

9

10 CFR 50.59. At the time of implementation of this modification, the regulation

provided that licensees may make changes to the facility or procedures as described in

the safety analysis report (SAR), without prior Commission approval, unless the

proposed change, test, or experiment involved a change to the Technical Specifications

incorporated in the license or an unreviewed safety question. Section 50.59(a)(2),

stated the following:

A proposed change, test, or experiment shall be deemed to involve an unreviewed

safety question: (i) if the probability of occurrence or the consequences of an accident

or malfunction of equipment important to safety previously evaluated in the safety

analysis report may be increased; (ii) if a possibility for an accident or malfunction of a

different type than any evaluated previously in the safety analysis report may be created;

or (iii) if the margin of safety as defined in the basis for any technical specification is

reduced.

At the time of the modification, any increase in probability of occurrence or consequence

was considered an unreviewed safety question. On October 4, 1999, the NRC issued a

revision to 10 CFR 50.59 in the Federal Register (64 FR 53582), effective 90 days after

approval of Regulatory Guide 1.187 (issued in November 2000). Among the changes

implemented to the revised rule, the NRC eliminated the term unreviewed safety

question, and clarified the requirements to allow changes, which involved only minimal

increases in probability or consequences to be made without prior NRC approval.

Because this performance deficiency did not meet the requirements of the revised rule

(which allowed for a minimal increase in frequency), it did not meet the criteria for

enforcement discretion, and therefore, was documented as a Severity Level IV, NCV,

consistent with the Enforcement Policy. On October 29, 2015, the NRC documented

this issue in NRC Inspection Report 05000458/2015007. The licensee entered the

performance deficiency into their corrective action program as Condition Report

CR-RBS-2015-7259 and did not deny the violation.

During the current inspection, the NRC team selected Condition Report

CR-RBS-2015-7259 as one of the samples reviewed to assess the adequacy of

the licensees problem identification and resolution program. The team found that

the licensee had not restored compliance with the rule and found several aspects

associated with how the licensee addressed the NCV that deviated from their corrective

action program as specified in Procedure EN-LI-102, Revision 25. These aspects

include:

The licensee initiated the condition report as significance C and directed it be

upgraded to a significance B ACE CARB [apparent cause evaluation, corrective

action review board] when the finding was issued as a NCV. The NRC documented

the finding on October 29, 2015, and yet the licensee did not upgrade the condition

report until December 17, 2015. This delayed initiation of the apparent cause

evaluation.

The licensee initially characterized the condition report as a significance C even

though it met two of the criteria listed in Attachment 9.1, Condition Report

Classification Guidance, of Procedure EN-LI-102 to be classified as significance B.

These examples included inadequate 10 CFR 50.59 review, evaluation or screening,

and Green NCV, Green finding violation, or traditional enforcement from the NRC.

10

The licensee closed the significance B condition report without having corrected the

condition adverse to quality (namely the 10 CFR 50.59 violation for failure to obtain

NRC approval prior to making a change to the facility). In order to restore

compliance, the licensee had three choices: (1) restore the facility to a condition that

did not require NRC approval (restore original design); (2) perform an adequate

evaluation that provided justification as to why the change did not increase the

probability or consequences of an accident by more than a minimal amount [based

upon the current standard] and deny the violation; or (3) submit a license

amendment request requesting NRC approval [after the fact] for a change to the

facility as described in the safety analysis report.

The licensee closed the condition report to a process that was not allowed by the

corrective action program. Section 5.5[5] CR [condition report] Disposition

Requirements of Procedure EN-LI-102 allows a condition report to be closed to

another condition report as long as the condition report being closed is the same as

or lower category level than the remaining condition report. Attachment 9.6 CR and

CA [corrective action] Closure to WMS [Work Management System] and Tracking,

also allows a condition report/corrective action to be closed to the work management

system if they have Condition Review Group approval. In this case, the licensee

closed the condition report to a licensing action request system, which was neither

another condition report nor part of the work management system, and therefore, not

allowed by the corrective action program. The licensees license action request

system did not have comparable controls or requirements for due date extensions as

specified by the corrective action program.

The corrective actions did not meet the guidance of Section 5.6[2] Corrective Action

Initiation of Procedure EN-LI-102 for crafting corrective actions, which states that

corrective action content should be specific, measurable, achievable, realistic, and

timely. The licensee did not initiate a corrective action to specifically address the

adverse condition. The action that was initiated was an indirect action that was

assigned to the Design Engineering department to provide technical input to the

Licensing department to support generation of a license amendment request for

submission to the NRC. Once the technical input was provided, the corrective action

and the condition report were closed. An adequate corrective action should have

required the Licensing department to obtain a license amendment accepting the

design change prior to closing the condition report. The failure to restore compliance

continues to exist up until the licensee receives a license amendment.

The closure review performed by the assigned manager failed to identify that the

condition adverse to quality had not been corrected and that the condition report was

not ready to close. Specific questions contained in Attachment 9.2, Checklist for

Level B CR Closure of Procedure EN-LI-102 that could have identified that the

condition report was not ready for closure include:

o Question 13 stated, verify the corrective actions corrected the condition

identified was checked SAT even though no corrective actions had been

generated to restore compliance.

11

o Question 14 stated, verify that each corrective action identified in the evaluation

and that was otherwise issued to address the condition was completed as

intended. Recommendations and enhancements may be tracked by other

processes, was checked SAT even though a sub question (also marked SAT)

stated, Verify the action item was not closed to a promise of a future action

item. In this case, the future action was implied that the Licensing department

would submit and obtain approval from the NRC for a license amendment.

o Questions 15 through 17 were left unchecked even though they were required to

be checked SAT, including Question 16 which states, verify the corrective

action is not closed to another process other than WO [work order] approved by

the CRG [Condition Review Group]. In this case the implied corrective action

(obtaining NRC approval) was closed to another process, which was not a work

order approved by the Condition Review Group.

o Question 19 (left blank) states, if this quality closure review identifies an

unsatisfactory closure of a checklist item annotated SAT, issue a corrective

action using the UNSAT RESPONSE PI" action type, with specific

recommendations or identified discrepancies that need further review. As noted

above, three questions that should have been annotated SAT, were left blank

and a corrective action was not generated using the UNSAT RESPONSE PI, as

required.

o Question 20 (left blank) states: When all items in the checklist are satisfactorily

completed, the CR is ready to close.

Corrective Action 14 was closed on October 6, 2016, even though the attached

closure checklist was not completed.

On April 12, 2017, the team determined that the licensee had not restored

compliance with this ongoing violation within a reasonable amount of time for

NCV 05000458/2015007-02, and that any future corrective actions could not be

considered timely and commensurate with the significance. The team concluded that

while the licensee originally met the criteria for dispositioning the 10 CFR 50.59 issue as

a NCV, based upon the fact that the condition report that documented the violation was

closed and the licensee had not restored compliance within a reasonable time (nearly

18 months), the team determined that the licensee no longer met the criteria for a NCV,

and therefore, this violation would be cited in a notice of violation.

Analysis. The team determined that the licensees failure to restore compliance within a

reasonable amount of time for a violation of regulatory requirements associated with a

design modification was a performance deficiency. Specifically, on October 29, 2015,

the NRC notified the licensee that a plant design change, which was implemented in

1999, had increased the probability of a loss-of-feedwater accident by more than a

minimal amount and was made without requesting prior NRC approval, was a violation

of 10 CFR 50.59 requirements. Title 10 CFR 50, Appendix B, Criterion XVI, requires in

part that, measures shall be established to assure that conditions adverse to quality,

such as failures, malfunctions, deficiencies, deviations, defective material and

equipment, and nonconformances are promptly identified and corrected. As of April 28,

2017, the licensee failed to correct a condition adverse to quality by restoring

compliance with 10 CFR 50.59.

12

The finding was more than minor because it is associated with the initiating events

aspect of the reactor safety cornerstone and affected the cornerstone objective to limit

the likelihood of events that upset plant stability and challenge critical safety functions

during power operations. Using Inspection Manual Chapter 0609, Appendix A, The

Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating

Events Screening Questions, dated June 19, 2012, the team determined that the

finding is of very low safety significance (Green). This was because the finding did not

cause a reactor trip and the loss of mitigation equipment relied upon to transition the

plant from the onset of the trip to a stable shutdown condition. The finding has a

human performance cross-cutting aspect associated with procedural adherence

because individuals failed to follow the procedures delineated by the corrective action

program [H.8].

The reactor oversight process (ROPs) significance determination process does not

specifically consider the regulatory process impact in its assessment of licensee

performance. Therefore, it is necessary to address this violation, which impedes the

NRCs ability to regulate, using traditional enforcement to deter non-compliance. Since

the violation is associated with a Green reactor oversight process violation, the

traditional enforcement violation was determined to be a Severity Level IV violation,

consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.

The NRCs Enforcement Policy dictates that severity level IV violations and violations

associated with green ROP findings are normally dispositioned as NCVs if they meet all

of the following: (1) the violation is placed into a corrective action program to restore

compliance and address recurrence; (2) the licensee must restore compliance within a

reasonable period of time (commensurate with the significance); (3) the violation must

either not be repetitive as a result of inadequate corrective action, or if repetitive, the

repetitive violation must not have been identified by the NRC (does not apply to green

ROP findings); and (4) the violation must not be willful. For the purposes of Criterion 2,

this includes actions taken to stop an ongoing violation from continuing (which should be

as soon as possible). The team concluded that while the licensee originally met the

criteria for dispositioning the issue 10 CFR 50.59 as a NCV; based upon the fact that the

licensee closed the condition report without restoring compliance, the licensee no longer

met the criteria for a NCV, and therefore, this violation will be cited in a notice of

violation.

Enforcement. The team identified a Severity Level IV, Green violation of 10 CFR 50.59,

Changes, Tests, and Experiments, Section (c)(2) which states in part that, a licensee

shall obtain a license amendment pursuant to Section 50.90 prior to implementing a

proposed change, test, or experiment if the change, test, or experiment would result in

more than a minimal increase in the frequency of occurrence of an accident previously

evaluated in the final safety analysis report (as updated). Contrary to the above, as of

April 28, 2017, the licensee failed to obtain a license amendment pursuant to

Section 50.90, prior to implementing a change, test, or experiment that resulted in a

more than minimal increase in the frequency of occurrence of an accident previously

evaluated in the final safety analysis report (as updated). Specifically, on July 3, 1999,

the licensee implemented a design change to the reactor core isolation cooling injection

location from the reactor vessel head to a feedwater line. However, the licensee failed

to correctly evaluate that a spurious reactor core isolation cooling actuation into the

feedwater line resulted in a more than minimal increase in the frequency of occurrence

13

of the loss of feedwater heating accident, previously evaluated in the updated final safety

analysis report. This performance deficiency was previously identified by the NRC and

documented as NCV 05000458/2015007-02. In accordance with Section 2.3.2.a of the

NRC Enforcement Policy, this finding is being cited because the licensee failed to

restore compliance within a reasonable amount of time after the violation was initially

identified. This finding was entered into the licensees corrective action program as

Condition Report CR-RBS-2017-03505, (VIO 05000458/2017009-01, Failure to Obtain

Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection Point.)

4OA6 Meetings, Including Exit

Exit Meeting Summary

On April 28, 2017, the inspectors presented the inspection results to Mr. Marvin Chase, Director,

Regulatory & Performance Improvement, and other members of the licensee staff. The licensee

acknowledged the issues presented. The licensee confirmed that any proprietary information

reviewed by the inspectors had been returned or destroyed.

14

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Chase, Director, Regulatory & Performance Improvement

A. Coates, Sr. Engineer, Regulatory Assurance

R. Crawford, Supervisor, Engineering

K. Huffstatler, Sr. Licensing Specialist, Regulatory Assurance

J. Lea, HVK System Engineer

P. Lucky, Manager, Performance Improvement

B. Maguire, Vice President, Operations

J. Reynolds, Sr. Manager, Operations

T. Schenk, Manager, Regulatory Assurance

K. Stupak, Manager, Training

T. Trask, Director, Recovery

S. Vazquez, Director, Engineering

NRC Personnel

J. Sowa, Senior Resident, River Bend Station

A. Vegel, Director, Division of Reactor Safety

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

Failure to Obtain Prior NRC Approval for a Change in Reactor

05000458/2017-009-01 NOV

Core Isolation Cooling Injection Point (Section 4OA2.5)

Discussed

Failure to Obtain Prior NRC Approval for a Change in Reactor

05000458/2015-007-02 NCV

Core Isolation Cooling Injection Point

LIST OF DOCUMENTS REVIEWED

Procedures

Number Title Revision

ADM-0073 Temporary Services and Equipment 307

AOP-0001 Reactor Scram 36

AOP-0029 Severe Weather Operation 38

CSP-0006 Chemistry Surveillance and Scheduling System 41

Attachment 1

Procedures

Number Title Revision

CSP-0109 Chemistry Surveillance for Non-Routing Samples 0

EN-LI-102 Corrective Action Program 24

EN-LI-102 Corrective Action Program 25

EN-LI-102 Corrective Action Program 26

EN-LI-102 Corrective Action Program 27

EN-LI-102 Corrective Action Program 28

EN-LI-102 Corrective Action Program 29

EN-LI-104 Self-Assessment and Benchmark Process 13

EN-LI-118 Cause Evaluation Process 21

EN-LI-118 Cause Evaluation Process 22

EN-LI-118 Cause Evaluation Process 23

EN-LI-118 Cause Evaluation Process 24

EN-LI-121 Trending and Performance Review Process 18

EN-LI-121 Trending and Performance Review Process 19

EN-LI-121 Trending and Performance Review Process 20

EN-LI-121 Trending and Performance Review Process 21

EN-LI-121 Trending and Performance Review Process 22

EN-OE-100 Operating Experience Program 27

EN-OP-104 Operability Determination Process 11

EN-QV-109 Audit Process 32

EN-RP-110-004 Radiation Protection Risk Assessment Process 7

EN-TQ-201 Systematic Approach to Training Process 22

GOP-001 Plant Startup 84

GOP-001 Plant Startup 85

GOP-001 Plant Startup 97

GOP-002 Power Decrease/Plant Shutdown 70

GOP-002 Power Decrease/Plant Shutdown 71

GOP-002 Power Decrease/Plant Shutdown 72

GOP-002 Power Decrease/Plant Shutdown 77

GOP-005 Power Maneuvering 321

A1-2

Procedures

Number Title Revision

GOP-005 Power Maneuvering 322

GOP-005 Power Maneuvering 323

GOP-005 Power Maneuvering 328

OSP-0014 Administrative control of Equipment and/or Devices 306

OSP-0022 Operations General Administration Guidelines 103

OSP-0043 Freeze Protection and Temperature Maintenance 30

SOP-0093 Condensate Demineralizer System 40

STP-000-0201 Monthly Operating Log 310

STP-000-0201 Monthly Operating Log 311

STP-251-3700 Fire System Yard Water Loop Flow Test 10

STP-251-3700 Fire System Yard Water Loop Flow Test 11

ML17160A401

Text

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