IR 05000298/2016001

NRC Integrated Inspection Report 05000298/2016001

ML16119A441
Person / Time
Site:	Cooper
Issue date:	04/28/2016
From:	Greg Warnick NRC/RGN-IV/DRP/RPB-C
To:	Limpias O Nebraska Public Power District (NPPD)
Warnick G
References
EA-15-089 IR 2016001
Download: ML16119A441 (45)

Text

April 28, 2016

SUBJECT:

COOPER NUCLEAR STATION

- NRC INTEGRATED INSPECTION REPORT 05000 298/2016001

Dear Mr. Limpias:

On March 31, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Cooper Nuclear Station.

On April 8, 2016, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented two findings of very low safety significance (Green) in this report. Both of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these NCVs, 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, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Cooper Nuclear Station

. If you disagree with a 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 Regional Administrator, Region IV; and the NRC resident inspector at the Cooper Nuclear Station.

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 NRC's Public 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/ Gregory G. Warnick , Chief Project Branch C Division of Reactor Projects Docket No. 50-298 License No

. DPR-46

Enclosure:

Inspection Report 05000298/2016001 w/ Attachment:

1. Supplemental Information 2. Request for Information for the O

REGION IV Docket: 05000 298 License: DPR-46 Report: 05000 298/20 16 0 01 Licensee: Nebraska Public Power District Facility: Cooper Nuclear Station Location: 72676 648A Ave Brownville, NE Dates: January 1 through March 31, 20 16 Inspectors:

P. Voss, Senior Resident Inspector C. Henderson, Resident Inspector W. Sifre, Senior Reactor Inspector M. Phalen, Senior Health Physicist J. O'Donnell, CHP, Health Physicist Approved By: Greg Warnick Chief, Project Branch C Division of Reactor Projects

- 2 -

SUMMARY

IR 05000 298/20 16 0 01; 01/01/20 16 - 03/31/20 16; Cooper Nuclear Station

Surveillance Testing

. The inspection activities described in this report were performed between January 1 and March 31, 2016 , by the resident inspectors at the Cooper Nuclear Station and inspectors from the NRC's Region IV office. Two findings of very low safety significance (Green) are documented in this report. Both of these findings involved violations of NRC requirements. 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."

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation of 10 CFR 50.55a, "Codes and Standards," for the licensee's failure to follow the ASME Code for Operation and Maintenance of Nuclear Power Plants when addressing the performance of reactor equipment cooling pump A within the high "required action range" of the inservice testing program. Specifically, on February 11, 2016, the licensee failed to follow ASME Subsection ISTB 6200(b) when engineering personnel, taking corrective action to address pump performance, failed to either correct the cause of the deviation or establish new reference values for the pump. Instead of establishing new reference values, the licensee performed an analysis to administratively raise the upper "required action range" limit, creating a wider range of acceptable pump operation than allowed by Table ISTB-5100-1, "Centrifugal Pump Test Acceptance Criteria." The licensee entered this issue into the corrective action program as Condition Report CR

-CNS-2016-00920, took action to reevaluate and rebaseline the pump with new reference values, and performed an extent of condition review to determine if other equipment was impacted by similar interpretations of the code.

The licensee's failure to establish new reference values for reactor equipment cooling pump A in accordance with the ASME Code was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the actions initially taken by the licensee would have required a relief request

could have delayed identification of a degrading pump trend due to t he creation of a wider range of acceptable operation; and the licensee's generic interpretation

, that the Table ISTB-5100-1 "acceptable range" could be administratively expanded , represented a programmatic vulnerability.

The inspectors used Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," and determined that the finding had very low safety significance (Green) because it did not represent a design or qualification deficiency, did not represent a loss of safety function for a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identificatio n and resolution associated with evaluation. Specifically, the licensee failed to thoroughly evaluate performance of reactor equipment cooling pump A in the "required action range" to ensure that the resolution correctly addressed the causes of the degraded performance

[P.2]. (Section 1R22)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensee's failure to follow Station Procedure 0.26, "Surveillance Program," and assess the operability of high pressure coolant injection steam line isolation instrumentation during surveillance testing. Specifically, the licensee failed to assess the operability of required isolation instrumentation when maintentance personnel opened terminal box 392 during surveillance testing and temporarily invalidated its environmental qualification. Licensee procedures required operations personnel to either establish compensatory measure s to restore the terminal box during an event , or declare the instrumentation inoperable and enter the applicable technical specification actions when the terminal box was opened. As an immediate corrective action

, the licensee implement ed Standing Order 2016-03, which directed operators to establish compensatory measures, if applicable, or declare the affected equipment inoperable when environmentally qualified terminal boxes would be opened during testing. The licensee entered this issue into their corrective action program for resolution as Condition Reports CR

-CNS-2016-00320 and CR

-CNS-2016-00476. The licensee's failure to assess the operability of high pressure coolant injection instrumentation when the associated terminal box was opened during surveillance testing, in violation of Station Procedure 0.26, was a performance deficiency.

The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, compo nent , and barrier performance attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to ensure the radiological barrier functionality of containment isolation. Specifically, with terminal box 392 open, its environmental qualification was temporarily invalidated, making the high pressure coolant injection low steam pressure and high steam flow containment isolation instrumentation inoperable during surveillance testing. In addition, two other terminal boxes and their associated surveillances were impacted by the performance deficiency. Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," the inspectors determined that the finding had very low safety significance (Green) because it: (1) did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components; and (2)did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect in the area of human performance associated with work management. Specifically, the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the identification and management of risk commensurate with opening terminal box 392 during surveillance testing

[H.5]. (Section 1R22)

PLANT STATUS

The Cooper Nuclear Station began the inspection period at full power. On February 12, 2016, the licensee lowered reactor power to approximately 70 percent in order to perform surveillance testing and planned work on reactor feedwater pump B. The plant returned to full power on February 13, 2016, where it remained for the rest of the reporting period, ex c e pt for minor reductions in power to support scheduled surveillance testing and rod pattern adjustments

. REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 04 Equipment Alignment (71111.04)

.1 Partial Walk

down

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk

-significant systems: January 17, 2016, High pressure coolant injection steam isolation instrumentation and control system for valves HPCI-MOV-15 and HPCI-MOV-16 February 19 , 2016, Service water cross connect valves SW

-MOV-36 and SW-MOV-37 and design flow requirement s February 29, 2016, Diesel generator sequential loading and kW loading analysis March 23, 2016, Instrument air system and reactor equipment cooling The inspectors reviewed the licensee's procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems were correctly aligned for the existing plant configuration

. These activities constitute d four partial system walkdown samples , as defined i n Inspection Procedure 71111.04.

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

On March 8, 2016, the inspectors performed a complete system walkdown inspection of the residual heat removal service water system. The inspectors reviewed the licensee's procedures and system design information to determine the correct system lineup for the existing plant configuration. The inspectors also reviewed outstanding work orders, open condition reports, in

-process design changes, temporary modifications, and other open items tracked by the licensee's operations and engineering departments

. The inspectors then visually verified that the system was correctly aligned for the existing plant configuration.

These activities constitute d one complete system walkdown sample , as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1 R 05 Fire Protection (71111.05)

Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensee's fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety

January 11, 2016, Reactor feed pump s area, Fire Area TB-A , Zon e 11E February 24 , 2016, Diesel generator room 1, Fire Area DG-A, Zone 14A and 14C February 24, 2016, Diesel generator room 2, Fire Area DG

-B, Zone 14B and 14D March 3 , 2016, Auxiliary relay room, Fire Area CB-D, Zone 8A For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensee's fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.

These activities constitute d four quarterly inspection sample s , as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1 R 06 Flood Protection Measures (71111.06)

a. Inspection Scope

On February 29, 2016, the inspectors completed an inspection of the station's ability to mitigate flooding due to internal causes. After reviewing the licensee's flooding analysis, the inspectors chose one plant area containing risk-significant structures, systems, and components that were susceptible to flooding:

Control building basement The inspectors reviewed plant design features and licensee procedures for coping with intern al flooding. The inspectors walked down the selected area to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.

These activities constitute d completion of one flood protection measures sample, as defined in Inspection Procedure 71111.06.

b. Findings

No findings were identified.

1 R 11 Licensed Operator Requalification Program and Licensed Operator Performance (71111.11)

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On February 19, 2016, the inspectors observed an evaluated simulator scenario performed by an operating crew. The inspectors assessed the performance of the operators and the evaluators' critique of their performance. The inspectors also assessed the modeling and performance of the simulator during the requalification activities

. These activities constitute d completion of one quarterly licensed operator requalification program sample , as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Review of Licensed Operator Performance

a. Inspection Scope

On February 13, 2016, the inspectors observed the performance of on-shift licensed operators in the plant's main control room. At the time of the observations, the plant was in a period of heightened activity due to maintenance and testing associated with a planned downpower to 70 percent. The inspectors observed the operators' performance of the following activities:

Rod manipulations associated with the downpower and rod pattern change , including the pre

-job brief Main steam isolation valve closure reactor protection system surveillance testing, including the pre-job brief In addition, the inspectors assessed the operators' adherence to plant procedures, including conduct of operations procedure and other operations department policies.

These activities constitute d completion of one quarterly licensed operator performance sample , a s defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1 R 12 Maintenance Effectiveness (71111.12)

a. Inspection Scope

The inspectors reviewed two instances of degraded performance or condition of safety-related or risk-significant structures, systems, and components (SSCs): February 29, 2016, Core spray reference leg injection January 19, 2016 , Reactor recirculation motor generator sets

The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensee's corrective actions. The inspectors reviewed the licensee's work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensee's characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.

These activities constitute d completion of two maintenance effectiveness samples , a s defined in Inspection Procedure 71111.12.

b. Findings

No findings were identified.

1 R 13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

The inspectors reviewed four risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:

February 3, 2016, Residual heat removal service water system maintenance window, Division II February 13, 2016, Feedwater pump B repair during planned downpower March 17, 2016, Reactor core isolation cooling maintenance window and northeast quad fan coil replacement March 25, 2016, Diesel generator system maintenance window, Division II The inspectors verified that these risk assessment s were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensee's risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments.

The inspectors also observed portions of two emergent work activities that had the potential to affect the functional capability of mitigating systems or to impact barrier integrity:

January 19, 2016, Diesel generator 1 repairs due to frequency starting time greater than technical specification surveillance requirements January 29, 2016, Loss of plant monitoring and information system/Gardel power supplies The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected structures, systems, and components.

These activities constitute d completion of six maintenance risk assessment and emergent work control inspection samples , a s defined in Inspection Procedure 71111.13.

b. Findings

No findings were identified.

1 R 15 Operability Determinations and Functionality Assessments (71111.15)

a. Inspection Scope

The inspectors reviewed five operability determinations that the licensee performed for degraded or nonconforming structures, systems, or components (SSCs):

January 15, 2016 , Operability determination of a control rod drive scram outlet valve due to leakage, CR-CNS-2016-0075 January 17, 2016, Operability determination of service water when the idle diesel generator is not isolated for a loss of offsite power/loss of coolant accident, CR-CNS-2016-00201 January 22, 2016 , Operability determination of the 125V Battery A due to lif ting positive battery posts, CR-CNS-2015-06703 March 16, 2016, Operability determination of reactor equipment cooling pump A due to exceedance of inservice testing required action limits, CR-CNS-2016-007 84 March 25, 2016, Operability determination of service water due to inconsistent pump column minimum wall thickness acceptance criteria, CR-CNS-2016-01448 The inspectors reviewed the timeliness and technical adequacy of the licensee's evaluations. Where the licensee determined the degraded SSC to be operable, t he inspectors verified that the licensee's compensatory measures were appropriate to provide reasonable assurance of operability. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded SSC

.

On January 25, 2016, the inspectors completed their review of operator actions taken or planned to compensate for degraded or nonconforming conditions.

The inspectors verified that the licensee effectively managed these operator workarounds to prevent adverse effects on the function of mitigating systems and to minimize their impact on the operator s' ability to implement abnormal and emergency operating procedures.

These activities constitute d completion of six operability and functionality review samples , which included one operator work

-around sample, as defined in Inspection Procedure 71111.15.

b. Findings

No findings were identified.

1 R 18 Plant Modifications (71111.18)

a. Inspection Scope

On March 9, 2016, the inspectors reviewed a permanent plant modification associated with the replacement of safety-related General Electric magne blast breakers with Siemens horizontal vacuum bottle circuit breakers which affected risk

-significant structures, systems, and components (SSCs)

.

The inspectors reviewed the design and implementation of the modification. The inspectors verified that work activities involved in implementing the modification did not adversely impact operator actions that may be required in response to an emergency or other unplanned event. The inspectors verified that post

-modification testing was adequate to establish the operability of the SSC s as modified.

These activities constitute d completion of one sample of permanent modifications , a s defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1 R 19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed six post-maintenance testing activities that affected risk

-significant structures, systems, or components (SSCs):

January 19, 2016, Diesel generator 1 emergent work February 3, 2016, Residual heat removal pump B relay and breaker maintenance February 3, 2016, Residual heat removal service water booster pump B discharge valve work February 3, 2016, Residual heat removal service water pump D outboard mechanical seal and discharge check valve inspection February 19, 2016, Residual heat removal and residual heat removal service water motor operated valve maintenance, Division II February 19, 2016, Torus to reactor vacuum breaker control switch replacement The inspectors reviewed licensing- and design

-basis documents for the SSCs and the maintenance and post-maintenance test procedures.

The inspectors observed the performance of the post

-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected SSCs

.

These activities constitute d completion of six post-maintenance testing inspection samples , a s defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1 R 22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors observed six risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the structures, systems, and components (SSCs) were capable of performing their safety functions:

In-service test s: January 25, 2016, High pressure coolant injection quarterly inservice test surveillance Other surveillance tests:

January 14 , 2016, High pressure coolant injection steam isolation valves HPCI-MOV-15 and HPCI-MOV-16 surveillance testing for primary containment isolation for the steam line break January 26, 2016, Service water quarterly and post

-loss of coolant accident flow surveillance acceptance criteria February 9, 2016, Diesel generator 31 day operability test, Division I February 19, 2016, Reactor equipment cooling pump A inservice testing surveillance March 3, 2016, SW

-MOV-36 and SW-MOV-37 surveillance testing The inspectors verified that these test s met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test s satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.

These activities constitute d completion of six surveillance testing inspection samples , a s defined in Inspection Procedure 71111.22.

b. Findings

(1) Failure to Follow ASME Code Requirements when taking Corrective Actions for a Pump in the Required Action Range Introduction

. The inspectors identified a Green, non-cited violation of 10 CFR 50.55a, "Codes and Standards," for the licensee's failure to follow the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM) when addressing the performance of reactor equipment cooling (REC) pump A within the high "required action range" of the inservice testing program. Specifically, the licensee failed to follow ASME Subsection ISTB-6200(b) when engineering personnel, while taking corrective action to address pump performance, failed to either correct the cause of the deviation or establish new reference values for the pump.

Description

. On February 11, 2016, the licensee performed Surveillance Procedure 6.1REC.101, "REC Surveillance Operation (IST)

- Div 1," for the two

-year comprehensive inservice test (IST) of REC pump A. This procedure was being used to meet the IST requirements for the pump in accordance with the 2001 Edition through the 200 3 Addenda of the ASME OM Code. During the IST, the pump exceeded the upper limit for discharge pressure required by the test, which put the pump in the high "required action range" in accordance with the ASME Code. Consistent with the surveillance procedure and ASME OM Subsection IST B-6200(b), the licensee declared the pump inoperable upon discovery of the condition. The licensee initiated condition report CR-CNS-2016-00784 to document the unacceptable inservice test results for REC pump A. ASME Section ISTB

-6200, "Corrective Action," Subsection (b), states, "If the measured test parameter values fall within the "required action range" of Table ISTB

-5100-1, the pump shall be declared inoperable until either the cause of the deviation has been determined and the condition is corrected, or an analysis of the pump is performed and new reference values are established in accordance with ISTB

-6200(c)." The licensee determined that the pump was operating acceptably, and as a result, determined that there was no need to correct the cause of the deviation in pump performance. However, the licensee did not establish new reference values, which serve to provide a baseline of acceptable pump performance, in accordance with ISTB

-6200(c). Instead, engineering personnel performed an analysis which assessed the operational readiness of the pump and evaluated pump performance trends as discussed in ISTB

-6200(c), but rather than rebaseline the pump, the licensee administratively raised the upper "required action" limit. Following this action, operations personnel declared REC pump A operable.

The inspectors reviewed the licensee's actions and challenged the site's decision to neither correct nor rebaseline the pump in accordance with Subsection ISTB

-6200(b). The inspectors observed that the licensee's actions to raise the upper "required action" limit inappropriately created a wider range of acceptable pump operation than allowed by ASME Table ISTB

-5100-1, "Centrifugal Pump Test Acceptance Criteria." In consultation with NRC regional and headquarters ASME Code experts, the inspectors concluded that these actions put the site in nonconformance with the ASME Code. The inspectors observed that the licensee's change would have required an NRC relief request and could have delayed identification of a degrading pump trend due to the creation of a wider range of acceptable operation. In response to inspector questions, the licensee determined that they had used this same method for different equipment on previous occasions. The inspectors determin ed that the licensee's generic interpretation that Table ISTB

-5100-1 acceptance criteria multipliers could be changed using Subsection ISTB-6200 represented a potential programmatic vulnerability.

This issue was entered into the licensee's corrective action program as CR

-CNS-201 6-00920, and the licensee subsequently took corrective action to establish new reference values for the pump

.

Analysis.

The licensee's failure to establish new reference values for REC pump A in accordance with the ASME OM code was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the actions initially taken by the licensee would have required a relief request

could have delayed identification of a degrading pump trend due to the creation of a wider range of acceptable operation; and the licensee's generic interpretation that the Table ISTB

-5100-1 "acceptable range" could be administratively expanded represented a programmatic vulnerability.

The inspectors used Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, and determined that the finding had very low safety significance (Green) because it did not represent a design or qualification deficiency, did not represent a loss of safety function for a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identificatio n and resolution associated with evaluation.

Specifically, the licensee failed to thoroughly evaluate performance of REC pump A in the "required action range" to ensure that the resolution correctly addressed the causes of the degraded performance

[P.2]. Enforcement

. Title 10 of the Code of Federal Regulations, Section 50.55a(b), "Codes and Standards," requires, in part, that systems and components of boiling and pressurized water cooled nuclear power reactors must meet the requirements of the ASME Code for Operation and Maintenance of Nuclear Power Plants.

Contrary to the above, on February 11, 2016, the licensee failed to ensure that systems and components in the plant met the requirements of the ASME OM Code. Specifically, the licensee failed to ensure ASME Subsection ISTB 6200(b) was met when engineering personnel, while taking corrective action to address REC pump A performance, failed to either correct the cause of the deviation or establish new reference values for the pump. Upon discovery, the licensee took action to reevaluate and rebaseline the pump with new reference values, and performed an extent of condition review to determine if other equipment was impacted by similar interpretations of the code. This violation is being treated as a non

-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy, because it was of very low safety significance (Green) and was entered into the licensee's corrective action program as Condition Report CR

-CNS-201 6-00920. (NCV 05000298/2016001

-01, "Failure to Follow ASME Code Requirements when taking Corrective Actions for a Pump in the Required Action Range

")

(2) Failure to Assess Operability of Technical Specification System Functions during Surveillance Testing

Introduction.

The inspectors identified a Green, non

-cited violation of Technical Specification (TS) 5.4.1.a, for the licensee's failure to follow Station Procedure 0.26, "Surveillance Program," and assess the operability of high pressure coolant injection (HPCI) steam line isolation instrumentation during surveillance testing. Specifically, the licensee failed to assess the operability of required HPCI isolation instrumentation when maintenance personnel opened terminal box (TB) 392 during surveillance testing and temporarily invalidated its environmental qualification.

Description.

On January 14, 2016, the inspectors observed surveillance testing for the Division 1 HPCI low steam pressure containment isolation function for HPCI

-MOV-15 in accordance with Station Procedure 6.1HPCI.701, "HPCI Steam Line Low Pressure Channel Functional Test (DIV 1)," Revision 5, under Work Order 5022860. During the surveillance test the inspectors identified that the licensee opened TB 392 to conduct the surveillance test

. This terminal box was identified as environmentally qualified (EQ). The inspectors questioned if TB 392 was EQ in the open condition during the surveillance test.

Following a review of documentation for the terminal box, the inspectors determined that TB 392 was only EQ in the closed condition in accordance with drawing CNS

-EQ-122, Sheet 1 and Sheet 2, "Cooper Nuclear Station EQ Configuration Detail Terminal Boxes and Equipment Enclosures," Revision 6 and Revision 5. Drawing CNS-EQ-122, Sheet 1 and Sheet 2, stated that the enclosure for TB 392 was credited with protecting terminal blocks from direct exposure to high

-energy line break (HELB) conditions and did not contain the field wires and Raychem splices that would allow the instrumentation to be EQ without an enclosure.

Station Procedure 0.26, "Surveillance Program," Revision 68 , Section 5 and Discussion Section 1.6 required the licensee to assess operability of TS system functions during surveillance testing

, and stated that delayed entry was only allowed if there was not a loss of function. Section 5 of this procedure stated, the "Shift Manager shall:

be aware of any other systems affected by the test and how they are affected

." Discussion Section 1.6 state d, "TS requirements may have notes that allow delayed entry into conditions and required actions for equipment made inoperable by performance of the surveillance. Even though delayed entry is allowed, the equipment/component is still considered inoperable while performing these surveillances. The delayed entry is only allowed if there is not a loss of function." Additionally, Station Procedure 0-Barrier, "Barrier Control Process," Revision 21, stated that opening terminal boxes in the reactor building required that either a compensatory measure be put in place or the SSC be declared inoperable.

The station did not implement a compensatory measure or declare instrumentation in the TB inoperable.

The inspectors and licensee concluded that the shift manager should have been aware of the impacts of opening the TB, and in accordance with procedures, should have declare d the TS system function inoperable for the HPCI low steam pressure and HPCI high steam fl ow isolation instrumentation when TB 392 was opened. Therefore, usage of the six hour delayed entry time for TS 3.3.6.1, "Primary Containment Isolation Instruments," was not allowed per Procedure 0.26 due to the instruments not being inoperable solely for surveillance testing.

As immediate corrective action s, the licensee identified additional TBs impacted by this concern, and implemented Standing Order 2016-03, which directed operators to either establish compensatory measures or declare the affected equipment inoperable when EQ TBs would be opened during testing. The licensee created long term corrective action s to assess whether compensatory measures could be justified for TBs opened during surveillance testing in the reactor building

, to assess whether open TBs could be qualified, and to update station procedures as required. The license entered this deficiency into their corrective action program for resolution as Condition Reports CR

-CNS-2016-00320 and CR-CNS-2016-00476.

Analysis.

The licensee's failure to assess the operability of HPCI isolation instrumentation when the associated terminal box was opened during surveillance testing, in violation of Station Procedure 0.26, was a performance deficiency.

The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the SSC and barrier performance attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to ensure the radiological barrier functionality of containment isolation. Specifically, with terminal box 392 open, its environmental qualification was temporarily invalidated, making the HPCI isolation instrumentation inoperable during surveillance testing. In addition, two other terminal boxes and their associated surveillances were impacted by the performance deficiency. Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: (1)did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components; and

(2) did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross

-cutting aspect in the area of human performance associated with work management. Specifically, the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the identification and management of risk commensurate with opening terminal box 392 during surveillance testing [H.5].

Enforcement.

Technical Specification 5.4.1.a, requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, "Quality Assurance Program Requirements," of February 1978. Section 1.f of Appendix A to Regulatory Guide 1.33 requires specific procedures for scheduling surveillance tests and calibration. The licensee established Station Procedure 0.26, "Surveillance Program," Revision 68, to schedule and control surveillance testing. Section 5 of Station Procedure 0.26 states, the "Shift Manager shall:

be aware of any other systems affected by the test and how they are affected." Contrary to the above, on January 14 , 2016, the licensee failed to ensure that the shift manager was aware of any other systems affected by the test and how they were affected during HPCI isolation surveillance testing.

Specifically, the licensee failed to assess the operability of all affected containment isolation instrumentation when maintenance personnel opened TB 392 during surveillance testing and temporarily invalidated its environmental qualification.

As immediate corrective action s , the licensee identified additional TBs impacted by the performance deficiency, and implemented Standing Order 2016-03, which directed operators to either establish compensatory measures or declare the affected equipment inoperable when environmentally qualified TB would be opened during testing.

This violation is being treated as a non

-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy, because it was of very low safety significance (Green) and was entered into the licensee's corrective action program as Condition Report s CR-CNS-201 6-0320 and CR-CNS-2016-00476. (NCV 05000298/2016001

-02, "Failure to Assess Operability of Technical Specification System Functions during Surveillance Testing")

Cornerstone: Emergency Preparedness

1 EP 6 Drill Evaluation (71114.06)

Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors observed an emergency preparedness drill on March 29, 2016, to verify the adequacy and capability of the licensee's assessment of drill performance. The inspectors reviewed the drill scenario, observed the drill from the Technical Support Center (TSC) and Simulator, and attended the post

-drill critique. The inspectors verified that the licensee's emergency classifications, off

-site notifications, and protective action recommendations were appropriate and timely. The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the licensee in the post-drill critique and entered into the corrective action program for resolution.

These activities constitute d completion of one emergency preparedness drill observation sample, as defined in Inspection Procedure 71114.06.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones:

Public Radiation Safety and Occupational Radiation Safety 2 RS 2 Occupational ALARA Planning and Controls (71124.02)

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors performed this portion of the attachment as a post

-outage review. During the inspection the inspectors interviewed licensee personnel, reviewed licensee documents, and evaluated licensee performance in the following areas:

Radiological work planning, including work activities of exposure significance, and radiological work planning ALARA evaluations, initial and revised exposure estimates, and exposure mitigation requirements.

The inspectors also verified that the licensee's planning identified appropriate dose reduction techniques, reviewed any inconsiste n cies between intended and actual work activity doses, and determined if post

-job (work activity) reviews were conducted to identify lessons learned. Specific work plans reviewed included refuel floor activities for the refuel bridge upgrades and radwaste processing for High

-Integrity Container (HIC) preparations for shipping.

Verification of dose estimates and exposure tracking systems including the basis for exposure estimates, and measures to track, trend, and if necessary reduce occupational doses for ongoing work activities. The inspectors evaluated the licensee's method for adjusting exposure estimates and reviewed the licensee's evaluations of inconsistent or incongruent results from the licensee's intended radiological outcomes.

Problem identification and resolution for ALARA planning and controls. The inspectors reviewed audits, self

-assessments, work

-in-progress and post

-job ALARA reviews, and corrective action program documents to verify problems were being identified and properly addressed for resolution.

These activities constitute d completion of two of the five required sample s of occupational ALARA planning and controls

, as defined in Inspection Procedure 71124.02.

b. Findings

No findings were identified.

2 RS 4 Occupational Dose Assessment (71124.04)

a. Inspection Scope

The inspectors evaluated the accuracy and operability of the licensee's personnel monitoring equipment, verified the accuracy and effectiveness of the licensee's methods for determining total effective dose equivalent, and verified that the licensee was appropriately monitoring occupational dose.

The inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas:

Source term characterization, including characterization of radiation types and energies, hard

-to-detect isotopes, and scaling factor s.

External dosimetry

, including National Voluntary Laboratory Accreditation Program (NVLAP) accreditation, storage, issue, use, and processing of active and passive dosimeters

.

The technical competency and adequacy of the licensee's internal dosimetry program.

Adequacy of the dosimetry program for special dosimetry situations, such as declared pregnant workers, multiple dosimetry placement, effective dose equivalent for external exposures (EDEX), shallow dose equivalent, neutron dose assessment, and dose records.

Problem identification and resolution for occupational dose assessment, including audits, self

-assessments, and corrective action documents.

These activities constitute d completion of five occupational dose assessment inspection samples, as defined in Inspection Procedure 71124.04.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security 4OA 1 Performance Indicator Verification (71151)

.1 Unplanned Scrams per 7000 Critical Hours (IE01)

a. Inspection Scope

The inspectors reviewed licensee event reports (LERs) for the period of January 1 through December 31, 2015 , to determine the number of scrams that occurred. The inspectors compared the number of scrams reported in these LERs to the number reported for the performance indicator. Additionally, the inspectors sampled monthly operating logs to verify the number of critical hours during the period.

The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the data reported.

These activities constituted verification of the unplanned scrams per 7000 critical hours performance indicator

, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Unplanned Power Changes per 7000 Critical Hours (IE03)

a. Inspection Scope

The inspectors reviewed operating logs, corrective action program records, and monthly operating reports for the period of January 1 through December 31, 2015, to determine the number of unplanned power changes that occurred. The inspectors compared the number of unplanned power changes documented to the number reported for the performance indicator. Additionally, the inspectors sampled monthly operating logs to verify the number of critical hours during the period. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the data reported.

These activities constituted verification of the unplanned power changes per 7000 critical hours performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

4OA 2 Problem Identification and Resolution (71152)

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensee's corrective action program and periodically attended the licensee's condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified.

The inspectors also reviewed the licensee's problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Annual Follow

-up of Selected Issues

a. Inspection Scope

The inspectors selected two issues for an in-depth follow

-up: On January 6, 2016, the inspectors reviewed entries in the control room log from the previous night shift, which discussed the identification of leakage into the scram discharge volume (SDV). Operations personnel had isolated the SDV in advance of performing planned maintenance on the system

, in order to quantify any potential leakage into the SDV, if it existed. During this activity, operators determined that there was no leakage into the North SDV, but the South SDV indicated leakage accumulating at a rate of 5.2 inches per hour. The inspectors noted that this was an indication of scram outlet valve leakage, and also observed that no condition report (CR) was written for the leakage that was discovered.

In response to inspector questions, operations personnel took action to initiate a CR (CR

-CNS-2016-00075) and assess operability. The inspectors noted that the site had failed to meet the requirements of Step 5.3.6.3 of Procedure 0-CNS-LI-102, "Corrective Action Process," which required, in part, that individuals ensure the condition was promptly documented on a Condition Report, by no later than the end of their shift.

In subsequent follow

-up with the licensee, the inspectors learned that the CR had been considered a non

-adverse condition, and as a result, CR generation had not been required.

The CR had been classified as a D

-trend non-adverse condition and closed. After further review, the inspectors determined that the condition met the licensee and NRC definition of a condition adverse to quality because the issue was a condition of an SSC, including failures and deficiencies, that could potentially render the SSC degraded or inoperable. Specifically, as discussed in GE SIL 173, "a leaking scram

[outlet] valve is of concern as the control rod drive (CRD) runs hot due to reactor water passing down through the drive and out the line to the scram discharge volume, and will continue to run hotter as the scram valve seat continues to erode. Eventually this could interfere with normal drive movement." In addition, the inspectors determined that scram outlet valve leakage into the SDV could result in high SDV water levels and undesirable scram signals if isolated, and could result in CRD drift if the leakage

bec ame excessive.

The inspectors determined that this issue represented a minor violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," which requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances. Licensee procedure 0-CNS-LI-102, "Corrective Action Process

," an Appendix B quality related procedure, provides instructions for identifying and classifying conditions adverse to quality. Procedure 0-CNS-LI-102, Attachment 1, states in part, that "adverse conditions are required to be corrected in the Corrective Action Program (CAP) and are subject to the rigor necessary to evaluate and thoroughly resolve important and significant issues." Contrary to the above, between January 6, 2016, and March 17, 2016, the licensee failed to assure that an adverse condition was corrected in the CAP and was subject to the rigor necessary to evaluate and thoroughly resolve important and significant issues. Specifically, the licensee initially failed to generate a condition report for indicated scram outlet valve leakage, and subsequently failed to classify the CR as a condition adverse to quality to ensure the deficiency would be resolved in the CAP. Instead, the CR was classified as D

-Trend, which denotes a non

-adverse condition that is handled outside of the CAP. The issue was minor in accordance with Inspection Manual Chapter 0612 Appendix B due to the minimal quantity of leakage identified and because other programmatic opportunities existed to identify the condition prior to significant plant impacts. Although this issue should be corrected, it constitutes a violation of minor significance that is not subject to enforcement action in accordance with Section 2 of the Enforcement Policy. The issue was entered into the licensee's CAP as CR-CNS-2016-01485. Licensee investigation revealed one CRD with slightly elevated temperatures, and the licensee generated a work order to repair the associated scram outlet valve.

The inspectors assessed the licensee's problem identification threshold and corrective actions to address the issue. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the condition

.

On January 11, 2016, the licensee identified that the Division 1 emergency diesel generator (EDG) was slow to start during a monthly surveillance test. Specifically, the EDG achieved rated voltage and frequency in 14.2 seconds, which exceeded the surveillance requirement limit of 14 seconds. The delayed start was linked to a degraded shuttle valve in the non

-safety portion of the air start system, which is normally bypassed during an emergency EDG start. The licensee determined that the apparent cause of the degradation was inadequate manufacturer controls of the component.

The inspectors assessed the licensee's problem identification threshold, cause analyses, and extent of condition reviews. The inspectors verified that the licensee appropriately prioritized the corrective actions and that these actions were adequate to correct the condition. These activities constitute d completion of two annual follow

-up samples as defined i n Inspection Procedure 71152.

b. Findings

No findings were identified.

4OA 5 Other Activities (Closed) Notice of Violation 05000298/2015007

-04, Failure to Evaluate the Lack of Missile Protection on the Emergency Diesel Generator 1 and 2 Fuel Oil Storage Tank Vents, EA-15-089 During the Component Design Basis Inspection conducted on April 6 through May 8, 2015, a violation of NRC regulations was identified and documented in NRC Inspection Report 05000298/2015007 (ML15173A450). The NRC had determined that a cited violation was associated with th e inspection.

The violation was cited because Cooper Nuclear Station (CNS) failed to restore compliance with NRC requirements within a reasonable amount of time after a previous violation was identified in NRC Inspection Report 05000298/2010007 (ML103370640

). In 2015, the team identified a Green, cited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," which states, in part, "Design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program." Specifically, the licensee failed to verify the adequacy of design of the vents for the emergency diesel generator (EDG) 1 and 2 fuel oil storage tanks to withstand impact from a tornado driven missile hazard, or to evaluate for exemption from missile protection requirements using an approved methodology.

The Notice of Violation (NOV) issued with the Inspection Report on June 22, 2015, required Cooper Nuclear Station to submit a written statement to the NRC within 30 days. The reply was required to contain the corrective steps taken to ensure full compliance was achieved. Cooper Nuclear Station submitted the response to the NRC on July 22, 2015 (ML15215A369). The corrective steps taken by the licensee included:

(1) incorporating a compilation of CNS and industry documentation into an engineering report to substantiate the conclusions of the design basis documents that evaluated the EDG fuel oil storage tank vents' ability to perform their design function following a design basis missile strike;

(2) removing the cap on the storage tank fill opening and installing a screen to ensure operability per the associated work order;

(3) reinforcing with engineers qualified to prepare or review calculations, the need to explicitly and literally state the technical issues when performing calculations; (4)incorporating lessons learned from the apparent cause evaluation as part of the Technical Rigor topic during engineering continuing training; and

(5) revising NEDC 13-046, Revision 1, to directly address all four tornado impact scenarios as described in Section XII, 2.3.3.2 of the CNS Updated Safety Analysis Report. The NRC responded in a letter to Cooper Nuclear Station's response on August 11, 2015 (ML15224B562). The letter stated that the NRC would inform the licensee if further inspection was warranted. The inspector reviewed the licensee's corrective actions associated with the violation. Specifically, the inspector reviewed Engineering Change EC-EE 15-012, "Diesel Generator Diesel Oil Tank Vents Torn ado Missile Analysis," Revision 1 , and Calculation NEDC 13-046, "Diesel Generator Storage Vent Line Tornado Missile Durability," Revision 2. Based on this review, the inspector concluded that the licensee had performed adequate corrective actions to restore compliance, address extent of condition, and prevent recurrence. No additional deficiencies were identified during the review of this Notice of Violation.

This review closes NOV 05000298/2015007

-04, "Failure to Evaluate the Lack of Missile Protection on the Emergency Diesel Generator 1 and 2 Fuel Oil Storage Tank Vents," EA-15-089. 4OA 6 Meetings, Including Exit

Exit Meeting Summary

On March 24, 2016, the inspectors presented the results of the diesel fuel oil tank Notice of Violation closure review to Mr. D. Buman, Director of Engineering, and other members of the licensee staff via telephone. The licensee acknowledged the inspection results. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On March 24 , 2016, the inspectors presented the radiation safety inspection results to Mr. K. Higginbotham, General Manager, Plant Operations, and other members of the licensee staff. The licensee acknowledged the inspection results. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On April 8, 2016, the inspectors presented the inspection results to Mr. O. Limpias, Vice President and Chief Nuclear Officer, 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.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

T. Barker, Manager, Engineering Program and Components

J. Bebb, Staff Health Physicist, Radiation Protection

D. Buman, Director, Engineering

B. Chapin, Manager, Maintenance

T. Chard, Manager, Quality Assurance

L. Dewhirst, Manager, Corrective Action and Assessment

K. Dia, Manager, System Engineering

J. Dixon, Supervisor, Radiation Protection

R. Estrada, Manager, Design Engineering

J. Flaherty, Senior Staff Engineer, Licensing

T. Forland, Engineer, Licensing

D. Goodman, Manager, Operations

K. Higginbotham, General Manager, Plant Operations

D. Kimball, Director, Nuclear

Oversight

O. Limpias, Vice President, Chief Nuclear Officer

J. Olberding, Licensing Engineer, Regulatory Affairs

R. Penfield, Director, Nuclear Safety Assurance

J. Shaw, Manager, Licensing

J. Stough, Manager, Emergency Preparedness

C. Sunderman, Manager, Radiation Protection

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000 298/2016001-01 NCV Failure to Follow ASME Code Requirements when taking Corrective Actions for a Pump in the Required Action Range

(Section 1R22)

05000298/2016001

-02 NCV Failure to Assess Operability of Technical Specification System Functions during Surveillance Testing (Section

1R22)

Closed

05000298/2015007

-04 VIO Failure to Evaluate the Lack of Missile Protection on the Emergency Diesel Generator 1 and 2 Fuel Oil Storage Tank Vents (Section 4OA5)