Text

NRC Integrated Inspection Report 05000416/2015004 and 07200050/2015001
	ML16043A104
Person / Time
Site:	Grand Gulf
Issue date:	02/11/2016
From:	Greg Warnick; NRC/RGN-IV/DRP/RPB-C
To:	Kevin Mulligan; Entergy Operations
	Warnick G
References
	IR 2015001, IR 2015004
	Download: ML16043A104 (75)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD.

ARLINGTON, TX 76011-4511 February 11, 2016 Kevin Mulligan Site Vice President Operations Entergy Operations, Inc.

Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 SUBJECT: GRAND GULF NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000416/2015004 AND 07200050/2015001

Dear Mr. Mulligan:

On December 31, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Grand Gulf Nuclear Station, Unit 1. On January 7, 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 three findings of very low safety significance (Green) in this report.

Three of these findings involved violations of NRC requirements. Additionally, NRC inspectors documented one Severity Level IV violation with no associated finding. Further, inspectors documented two licensee-identified violations that were determined to be of very low safety significance (Green) in this report. The NRC is treating these violations as non-cited violations consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these non-cited violations, 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 Grand Gulf 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 Grand Gulf 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 NRCs 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/

Greg Warnick, Branch Chief Project Branch C Division of Reactor Projects Docket No. 50-416;72-050 License No. NPF-29

Enclosure:

Inspection Report 05000416/2015004 and 07200050/2015001 w/ Attachments:

1. Supplemental Information 2. Request for Information - O

REGION IV==

Docket: 05000416 and 07200050 License: NPF-29 Report: 05000416/2015004 and 07200050/2015001 Licensee: Entergy Operations, Inc.

Facility: Grand Gulf Nuclear Station, Unit 1 Location: 7003 Baldhill Road Port Gibson, MS 39150 Dates: October 1 through December 31, 2015 Inspectors: M. Young, Senior Resident Inspector N. Day, Resident Inspector D. Loveless, Senior Reactor Analyst J. Buchanan, Physical Security Inspector H. Freeman, Senior Reactor Inspector N. Greene, PhD, Health Physicist G. Guerra, CHP, Emergency Preparedness Inspector M. Phalen, Senior Health Physicist G. Pick, Senior Reactor Inspector E. Simpson, ISFSI Inspector Approved Greg Warnick, Chief, Project Branch C By: Division of Reactor Projects-1- Enclosure

SUMMARY

IR 05000416/2015004, 07200050/2015001; 10/01/2015 - 12/31/2015; Grand Gulf Nuclear

Station; Maintenance Effectiveness, Maintenance Risk Assessments and Emergent Work Control, Operability Determinations and Functionality Assessments, and Post-Maintenance Testing The inspection activities described in this report were performed between October 1 and December 31, 2015, by the resident inspectors at the Grand Gulf Nuclear Station and inspectors from the NRCs Region IV office. Three findings of very low safety significance (Green) are documented in this report. Three of these findings involved violations of NRC requirements. Additionally, NRC inspectors documented one Severity Level IV violation with no associated finding. Further, inspectors documented two licensee-identified violations that were determined to be of very low safety significance (Green) in this report. 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 NRCs 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 reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, for the failure to establish adequate instructions to perform a simulated surveillance on the division I diesel generator. Specifically, the simulated surveillance run instructions verified the trip high vibration (E-23H) valve was open, but it did not close the (E-23H) valve following the run to ensure the high vibration trip was bypassed. As a result, the division I diesel generator spuriously tripped on high vibrations during the November 21, 2015, run and was rendered inoperable and unavailable. On November 22, 2015, the licensee closed the trip high vibration (E-23H) valve and successfully ran the division I diesel generator to return it to operable status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-6831.

The failure to establish adequate preventative maintenance instructions to perform a division I diesel generator simulated run and return the valve lineup to the required position was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment 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, following the division I diesel generator simulated run, the preventative maintenance instruction did not require the licensee to close the trip high vibration (E-23H) valve, and therefore the high vibration trip capability remained for a duration of approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . As a result, during the November 21, 2015 run, the diesel generator spuriously tripped on an invalid high vibration signal and was rendered inoperable and unavailable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it:

(1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program.

The inspectors determined that the finding has a design margin cross-cutting aspect within the human performance area because the licensee failed to ensure margins are carefully guarded and changed only through a systematic and rigorous process. Specifically, the licensee failed to fully implement their design change process such that all effected station documents and procedures were identified and revised after removing the high vibration trip for the division I and division II diesel generators [H.6]. (Section 1R12)

Green.

The inspectors identified a non-cited violation of Technical Specification Surveillance Requirement 3.0.1, for the failure to follow requirements when a surveillance was not performed within the specified frequency and declare the Limiting Condition for Operation not met or follow the provisions in Surveillance Requirement 3.0.3. Specifically, the licensee did not follow Technical Specification Surveillance Requirement 3.0.1, when they discovered that Surveillance Requirement 3.8.1.9 was not performed within its specified frequency and either declare Technical Specification Limiting Condition for Operation 3.8.1 not met, or perform the required actions to determine whether compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed. The licensee failed to enter Technical Specification Surveillance Requirement 3.0.1, until September 29, 2015, after discussions with the NRC. On September 29, 2015, the licensee adequately performed the actions required in Technical Surveillance Requirement 3.0.3. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-5602.

The failure to timely enter and perform the actions as required per Technical Specification Surveillance Requirement 3.0.1 was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment 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 failure to perform technical specification surveillance requirements, and associated actions, did not ensure that the diesel generator could appropriately respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2,

Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program.

The inspectors determined that the finding has a conservative bias cross-cutting aspect within the human performance area because the licensee failed to use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, operations personnel failed to enter Technical Specification Surveillance Requirement 3.0.1 because the operability determination alone justified operability without doing a detailed risk evaluation [H.14]. (Section 1R13)

Green.

The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, for the failure to establish adequate maintenance instructions to perform work activities on the division III diesel generator overspeed trip limit switch.

Specifically, work orders did not contain adequate instructions to check the overspeed trip switches alignment in accordance with vendor recommendations. As a result, the division III diesel generator was rendered inoperable and unavailable. On July 15, 2015, the licensee appropriately set the limit switch to overspeed actuating arm engagement, and returned the diesel generator to operable. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-3985.

The failure to establish adequate work instructions to verify the overspeed switch was properly set and adjusted was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment 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, work orders to check the overspeed trip switches alignment did not contain adequate instructions to successfully perform the maintenance. The division III diesel generator was declared inoperable when the diesel spuriously tripped during the monthly surveillance run on July 13, 2015. The inspectors performed the initial significance determination for the division III emergency diesel generator failure. The inspectors used the NRC Inspection Manual 0609,

Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The finding required a detailed risk evaluation because it involved a performance deficiency that represented a loss of the high pressure core spray system following a postulated loss of offsite power because of the failure of the division III diesel generator. The Region IV senior reactor analyst performed a detailed risk evaluation in accordance with NRC Inspection Manual 0609,

Appendix A, Section 6.0, Detailed Risk Evaluation. The detailed risk evaluation result is a finding of very low safety significance (Green). The calculated change in core damage frequency of 5.0 x 10-7 was dominated by an unrecovered station blackout beyond battery depletion. The analyst determined that the bounding risk of a large, early release of radiation was 9.6 x 10-8. For the details of the analysis, see Attachment 3.

Work orders were developed to address operating experience provided from the diesel generator vendor to the industry in December 2011. The inspectors determined that the cause of the deficiency occurred in 2011, and therefore, determined the finding did not have a cross-cutting aspect since it is not indicative of current licensee performance.

(Section 1R19)

Cornerstone: Barrier Integrity

SLIV. The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.72(b)(3)(v)(C), for the licensees failure to make a required eight-hour report to the NRC for a condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

Specifically, on October 14, 2015, the licensee failed to make the required eight-hour report following two primary containment isolation valves, 1P11F130 and 1P11F131, in the same flow path being declared inoperable. On October 15, 2015 at 9:07 pm, the licensee made a late Event Notification, EN 51473. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-6043.

The failure to make an eight-hour report, as required by 10 CFR 50.72(b)(3)(v)(C), for a condition that could have prevented fulfillment of a safety function was a performance deficiency. This performance deficiency was screened using Inspection Manual Chapter 0612 and was determined to be a minor violation in the Reactor Oversight Process.

However, due to the performance deficiency affecting the NRCs ability to perform its regulatory oversight function, this performance deficiency was evaluated for traditional enforcement in accordance with the NRC Enforcement Policy. This performance deficiency was determined to be a Severity Level IV violation in accordance with Section 6.9.d.9 of the NRC Enforcement Policy, dated February 4, 2015. No cross-cutting aspect was assigned to this violation because no Reactor Oversight Process finding existed. (Section 1R15)

Licensee-Identified Violations

Two violations of very low safety significance (Green) that were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and their associated corrective action tracking numbers are listed in Section 4OA7 of this report.

PLANT STATUS

The Grand Gulf Nuclear Station began the inspection period at 100 percent power.

On October 8, 2015, the operators reduced power to approximately 85 percent to perform partial rod exercises and pattern adjustment. Upon completion, operators performed power ascension activities to reach 100 percent power on October 9, 2015.

From November 12 - 23, 2015, the operators reduced power to approximately 53 percent to perform control rod sequence exchange, settle time testing, partial rod exercises and pattern adjustments. Upon completion, operators performed power ascension activities to reach 100 percent power on November 23, 2015.

On December 10, 2015, the operators reduced power to approximately 81 percent to perform partial rod exercises. Upon completion, operators performed power ascension activities to reach 100 percent power on December 12, 2015.

On December 18, 2015, the operators reduced power to approximately 83 percent to perform partial rod exercises. Upon completion, operators performed power ascension activities to reach 99 percent power on December 19, 2015.

From December 28 - 31, 2015, the operators reduced power to approximately 59 percent to perform power suppression testing. Upon completion, operators performed power ascension activities to reach 87 percent power on December 31, 2015.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Main article: IP 71111.01

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On November 17, 2015, the inspectors completed an inspection of the stations readiness for impending adverse weather conditions. The inspectors reviewed plant design features, the licensees procedures to respond to tornadoes and high winds, and the licensees implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

These activities constituted one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

.2 Readiness to Cope with External Flooding

a. Inspection Scope

On November 17, 2015, the inspectors completed an inspection of the stations readiness to cope with external flooding. After reviewing the licensees flooding analysis, the inspectors chose three plant areas that were susceptible to flooding:

diesel generator building and associated flood barrier doors

control building and associated flood barrier doors

control building and auxiliary building roofs The inspectors reviewed plant design features and licensee procedures for coping with flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether credited operator actions could be successfully accomplished.

These activities constituted one sample of readiness to cope with external flooding, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

1R04 Equipment Alignment

Main article: IP 71111.04

.1 Partial Walkdown

a. Inspection Scope

The inspectors performed partial system walk-downs of the following risk-significant systems:

October, 21, 2015, division II diesel generator while the division I diesel generator was in maintenance

December 9, 2015, residual heat removal A while residual heat removal B was inoperable for remote shutdown test

December 10, 2015, standby service water A while division II diesel generator was inoperable

December 10, 2015, standby service water C while division II diesel generator was inoperable The inspectors reviewed the licensees 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 constituted four partial system walk-down samples, as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

Main article: IP 71111.05

.1 Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on five plant areas important to safety:

October 16, 2015, fire area 64, fire zones 1M110 and 1M112, standby service water pump house A and valve room

October 16, 2015, fire area 65, fire zones 2M110 and 2M112, standby service water pump house B and valve room

October 16, 2015, fire area 30, fire zone 0C214, division I switchgear area (Unit 2)

November 19, 2015, fire areas 20, fire zone 1A407, division II motor control center 16B41 room

November 19, 2015, fire areas 21, fire zone 1A410, division I motor control center 15B21 room For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees 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 constituted five quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

Main article: IP 71111.06

a. Inspection Scope

On December 10, 2015, the inspectors completed an inspection of the stations ability to mitigate flooding due to internal causes. After reviewing the licensees flooding analysis, the inspectors chose two plant areas containing risk-significant structures, systems, and components that were susceptible to flooding:

high pressure core spray pump room

residual heat removal C pump room The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas 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 constituted completion of two flood protection measures samples, as defined in Inspection Procedure 71111.06.

b. Findings

No findings were identified.

1R07 Heat Sink Performance

Main article: IP 71111.07

a. Inspection Scope

On October 20 - 21, 2015, the inspectors completed an inspection of the readiness and availability of risk-significant heat exchangers. The inspectors observed the licensees inspection of the division I diesel generator jacket water cooling heat exchanger and the material condition of the heat exchanger internals. Additionally, the inspectors walked down the heat exchanger to observe its performance and material condition, reviewed tube plugging data sheets and associated performance calculations, and verified that the heat exchanger was correctly categorized under the Maintenance Rule and was receiving the required maintenance.

These activities constituted completion of one heat sink performance annual review sample, as defined in Inspection Procedure 71111.07.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

Main article: IP 71111.11

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On November 16, 2015, the inspectors observed simulator training for an operating crew. The operating crew completed a training scenario that required operation of the plant in the MELLA+ operating region. 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 training activity.

These activities constituted 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 November 12 - 13, 2015, the inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened activity due to a downpower to 55 percent for control rod sequence exchange, settle time testing, and monthly operability checks for control rod withdrawal blocks. The inspectors observed the operators performance of communications during the downpower, procedural adherence during control rod manipulation, and interaction between operators and reactor engineering.

In addition, the inspectors assessed the operators adherence to plant procedures, including procedure EN-OP-115, Conduct of Operations, Revision 15, and other operations department policies.

These activities constituted completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

a. Inspection Scope

The inspectors reviewed two instances of degraded performance or condition of safety-related structures, systems, and components (SSCs):

December 23, 2015, standby liquid control system due to loss of continuity on the standby liquid control B squib valve

December 30, 2015, division I diesel generator due to a high vibration trip during a surveillance test The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensees 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 constituted completion of two maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.

b. Findings

Introduction.

The inspectors reviewed a Green, self-revealing, non-cited violation of Technical Specification 5.4.1.a, for the failure to establish adequate instructions to perform a simulated surveillance on the division I diesel generator. Specifically, the simulated surveillance run instructions verified the trip high vibration (E-23H) valve was open, but it did not close the (E-23H) valve following the run to ensure the high vibration trip was bypassed. As a result, the division I diesel generator spuriously tripped on high vibrations during the November 21, 2015, run and was rendered inoperable and unavailable.

Description.

On November 21, 2015, the licensee was performing a monthly surveillance test on the division I diesel generator. This diesel generator run was also categorized as a post maintenance run to ensure that work done on the air start, lube oil, and voltage regulator systems was appropriate and correct. During this run, the diesel generator spuriously tripped on high vibrations.

Engineering Change 51435 was completed August 27, 2014. The reason for the change was to bypass the vibration trip system for the division I and division II diesel generators.

The purpose of the vibration trip was to provide equipment protection should high engine vibrations occur. However, industry experience indicated that the switches are unreliable in performing that function, causing spurious and unwanted trips, and subsequent system unavailability. Therefore, the licensee determined that it was acceptable to bypass and/or disable the engine vibration trips by closing and administratively locking the manual isolation valve (E-23H). This engineering change was developed and implemented via Corrective Action 11 of Condition Report CR-GGN-2013-5899 after a similar non-valid high vibration diesel generator trip occurred on the division I diesel generator.

Before the division I diesel run took place, the licensee performed a simulated run, in accordance with Preventative Maintenance Instruction 07-S-23-P75-3, Div I and Div II Diesel Generator Simulated Run, Revision 7, to ensure that the pneumatic computer logic board was appropriately set, following maintenance on the air start, lube oil, and voltage regulator systems. This procedure was last revised on March 25, 2008. During this simulated run, the procedure required verification that the TRIP HIGH VIBRATION E-23H valve is open, but it does not require the valve to be closed after the run. As written, the procedure restored and maintained the high vibration diesel generator trip.

Following the successful simulated run, on November 21, 2015, the licensee ran the division I diesel generator. Shortly after reaching full load, the diesel generator spuriously tripped on high engine vibrations and was declared inoperable and unavailable. On November 22, 2015, the licensee positioned the E-23H valve closed, successfully ran an operability test of the division I diesel generator with additional vibration monitoring, and declared it operable. Therefore, the high vibration trip vulnerability on division I diesel generator existed for approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . The licensee entered this into their corrective action program as Condition Report CR-GGN-2015-6831.

Analysis.

The failure to establish adequate preventative maintenance instructions to perform a division I diesel generator simulated run and return the valve lineup to required position was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment 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, following the division I diesel generator simulated run, the preventative maintenance instruction did not require the licensee to close the trip high vibration (E-23H) valve, and therefore the high vibration trip capability remained for a duration of approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . As a result, during the November 21, 2015 run, the diesel generator spuriously tripped on an invalid high vibration signal and was rendered inoperable and unavailable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it:

(1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of functionality;

(2) did not represent a loss of system and/or function; (3)did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and

(4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program.

The inspectors determined that the finding has a design margin cross-cutting aspect within the human performance area because the licensee failed to ensure margins are carefully guarded and changed only through a systematic and rigorous process.

Specifically, the licensee failed to fully implement their design change process such that all effected station documents and procedures were identified and revised after removing the high vibration trip for the division I and division II diesel generators [H.6].

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 of Regulatory Guide 1.33, Revision 2. Section 9.a of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for performing maintenance, such that, maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with documented instructions appropriate to the circumstances. The licensee established Preventative Maintenance Instruction, 07-S-23-P75-3, Div I and Div II Diesel Generator Simulated Run, Revision 7, to meet the Regulatory Guide 1.33 requirement. Contrary to the above, on November 21, 2015, the licensee failed to establish documented instructions appropriate to the circumstances. Specifically, the licensee used Preventative Maintenance Instruction, 07-S-23-P75-3, Div I and Div II Diesel Generator Simulated Run, Revision 7, to perform a simulated diesel generator run but did not ensure the high vibration trip was bypassed before the instructions were concluded. As a result, during the November 21, 2015, division I diesel generator run, the diesel spuriously tripped on an invalid high vibration trip signal. On November 22, 2015, the licensee closed the trip high vibration (E-23H) valve and successfully ran the division I diesel generator to return it to operable status. Because this finding is determined to be of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-GGN-2015-6831, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000416/2015004-01, Failure to Have Appropriate Instructions for Preventative Maintenance on the Division I Diesel Generator Simulated Run.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

a. Inspection Scope

On October 1, 2015, the inspectors reviewed a risk assessment performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk required by Technical Specification Surveillance Requirement (SR) 3.0.3 for failure to perform SR 3.8.1.9 on division I, II, and III diesel generators.

The inspectors verified that this risk assessment was 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 licensees risk assessment and verified that the licensee implemented appropriate risk management actions based on the result of the assessment.

Additionally, on October 1, 2015, the inspectors observed portions of one emergent work activity, secondary containment door seal replacements, after the failure of the secondary containment drawdown surveillance test that had the potential to impact barrier integrity.

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 (SSCs).

These activities constituted completion of two maintenance risk assessments and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.

b. Findings

Introduction.

The inspectors identified a Green, non-cited violation of Technical Specification SR 3.0.1, for the failure to follow requirements when a surveillance was not performed within the specified frequency and declare the Limiting Condition for Operation (LCO) not met or follow the provisions in SR 3.0.3. Specifically, the licensee did not follow Technical Specification SR 3.0.1, when they discovered that SR 3.8.1.9 was not performed within its specified frequency and either declare Technical Specification LCO 3.8.1 not met, or perform the required actions to determine whether compliance with the requirement to declare the LCO not met may be delayed.

Description.

Technical Specification SR 3.8.1.9 states, Verify each DG rejects a load greater than or equal to its associated single largest post-accident load and engine speed is maintained less than nominal plus 75 percent of the difference between nominal speed and the overspeed setpoint of 15 percent above nominal, whichever is lower.

However, the technical specification surveillance requirements were not fulfilled during the testing. This was identified during the Grand Gulf 2015 Component and Design Basis Inspection and dispositioned as a Green, non-cited violation of Technical Specification 3.8.1 (NCV 050000416/2015007-06, Failure to Perform Surveillance Requirement 3.8.1.9.)

The licensee determined that this was a missed surveillance on August 15, 2015.

However, the licensee was able to provide reasonable expectation that the emergency diesel generators were capable of a largest load reject, by having successfully completed Technical Specification SR 3.8.1.10, which ensured that each diesel generator was able to reject a load greater than its respective single largest load.

However, the NRC determined that Technical Specification SR 3.8.1.10 had never been performed, based on SR 3.0.1 and guidance from Inspection Manual Chapter 0326.

Per Attachment 2 of NRC Inspection Manual Chapter 0326, SR 3.0.3 may not be applied when a licensee discovers that a technical specification surveillance has never been performed. In cases where a specified safety function or a necessary and related support function required for operability has never been performed, then a reasonable expectation of operability does not exist. However, Technical Specification SR 3.0.3 would apply should the licensee determine that a technical specification surveillance had been demonstrated outside of routine surveillances, e.g., for post-maintenance testing, or for testing resulting from normal or off-normal plant operations.

Since the licensee was able to justify that Technical Specification SR 3.8.1.10 bounded Technical Specification SR 3.8.1.9, the use of the guidance in Inspection Manual Chapter 0326 for the scenario of a missed surveillance was appropriate.

The licensee failed to enter SR 3.0.3 until September 29, 2015, when the inspectors asked for the SR 3.0.3 required risk assessment. At that point, the licensee adequately performed the actions required in SR 3.0.3. The licensee entered this into their corrective action program as Condition Report CR-GGN-2015-5602.

Analysis.

The failure to timely enter and perform the actions as required per Technical Specification SR 3.0.1 was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment 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 failure to perform technical specification surveillance requirements, and associated actions, did not ensure that the diesel generator could appropriately respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it:

(1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of functionality;

(2) did not represent a loss of system and/or function;

(3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and

(4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program.

The inspectors determined that the finding has a conservative bias cross-cutting aspect within the human performance area because the licensee failed to use decision making-practices that emphasize prudent choices over those that are simply allowable.

Specifically, operations personnel failed to enter Technical Specification SR 3.0.1 because the operability determination alone justified operability without doing a detailed risk evaluation [H.14].

Enforcement.

Technical Specification SR 3.0.1, states, in part, that the failure to perform a surveillance within the specified frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Technical Specification SR 3.0.3 provided actions, such that, compliance with the requirement to declare the LCO not met may be delayed. Contrary to the above, on August 15, 2015, the licensee did not follow SR 3.0.1 when a surveillance was not performed within the specified frequency and declare the LCO not met or follow the provisions in SR 3.0.3. Specifically, the licensee did not follow SR 3.0.1, when they discovered that SR 3.8.1.9 was not performed within its specified frequency and either declare Technical Specification LCO 3.8.1 not met, or perform the required actions to determine whether compliance with the requirement to declare the LCO not met may be delayed. The licensee failed to enter SR 3.0.1, until September 29, 2015, after discussions with the NRC. On September 29, 2015, the licensee adequately performed the actions required in SR 3.0.3. Because this finding is determined to be of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-GGN-2015-5602, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000416/2015004-02, Failure to Timely Enter Technical Specification Surveillance Requirement 3.0.1.

1R15 Operability Determinations and Functionality Assessments

Main article: IP 71111.15

a. Inspection Scope

The inspectors reviewed one operability determination and one functionality assessment that the licensee performed for degraded or nonconforming structures, systems, or components (SSCs):

October 14 - 15, 2015, operability determination for primary containment isolation valves 1P11F130, 1P11F131, and 1E12F346 because the required local leak rate test was not performed at the required post extended power uprate pressure

October 20 - 23, 2015, functionality assessment of Claiborne County emergency sirens following failures on July 28, 2015, and July 31, 2015 The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable or functional, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability or functionality. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability or functionality of the degraded SSC.

These activities constituted completion of two operability and functionality review samples, as defined in Inspection Procedure 71111.15.

b. Findings

Introduction.

The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.72(b)(3)(v)(C), for the licensees failure to make a required eight-hour report to the NRC for a condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

Specifically, on October 14, 2015, the licensee failed to make the required eight-hour report following two primary containment isolation valves, 1P11F130 and 1P11F131, in the same flow path being declared inoperable.

Description.

On October 14, 2015, at 12:20 pm, the licensee identified that there were two primary containment isolation valves, 1P11F130 and 1P11F131, in the same flow path that were not local leak rate tested using the post-extended power uprated peak containment pressure. The licensee declared both valves inoperable and closed a valve in the flow path to restore leakage to within limits in the completion time of four hours as stated in Technical Specification 3.6.1.3, Condition C. The licensee made the determination that the condition was not a reportable event.

On October 15, 2015, at 8:00 am, the inspectors further questioned the licensee about the condition of the penetration. While investigating the NRCs questions, the licensee performed a re-evaluation for reportability. Subsequently, the licensee determined that the two valves were in the same penetration and both were declared inoperable, therefore this condition was considered a potential loss of safety function of a single train system that is needed to control radiation release.

On October 15, 2015, at 9:07 pm, the licensee initiated an eight-hour report, EN51473, to the NRC for a condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. This report was approximately 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months after the time of discovery.

Analysis.

The failure to make an eight-hour report, as required by 10 CFR 50.72(b)(3)(v)(C), for a condition that could have prevented fulfillment of a safety function was a performance deficiency. This performance deficiency was screened using Inspection Manual Chapter 0612 and was determined to be a minor violation in the Reactor Oversight Process. However, due to the performance deficiency affecting the NRCs ability to perform its regulatory oversight function, this performance deficiency was evaluated for traditional enforcement in accordance with the NRC Enforcement Policy. This performance deficiency was determined to be a Severity Level IV violation in accordance with Section 6.9.d.9 of the NRC Enforcement Policy, dated February 4, 2015. No cross-cutting aspect was assigned to this violation because no Reactor Oversight Process finding existed.

Enforcement.

Title 10 of the Code of Federal Regulations Part 50.72(b)(3)(v)(C),requires, in part, that licensee shall notify the NRC within eight hours of the occurrence of an event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above, on October 14, 2015, the licensee failed to notify the NRC within eight hours of the occurrence of an event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Specifically, on October 14, 2015, the licensee failed to make the required eight-hour report following two primary containment isolation valves, 1P11F130 and 1P11F131, in the same flow path being declared inoperable. On October 15, 2015, at 9:07 pm, the licensee made a late Event Notification, EN 51473. Because this violation has been entered into the licensees corrective action program as Condition Report CR-GGN-2015-6043, safety function was restored within a reasonable time, and the violation was not repetitive or willful, this Severity Level IV violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000416/2015004-03, Failure to Make a Required Eight-Hour Report for Loss of Safety Function.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

a. Inspection Scope

The inspectors reviewed four post-maintenance testing activities that affected risk-significant structures, systems, or components (SSCs):

October 23, 2015, division I diesel generator following an extended maintenance outage

November 27, 2015, high pressure core spray system following jockey pump seal replacement and leaking fitting repair

December 22, 2015, standby service water train B, fan D following preventative maintenance

December 30, 2015, division I emergency switchgear and battery room ventilation heater following temperature 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 constituted completion of four post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

b. Findings

Introduction.

The inspectors reviewed a Green, self-revealing, non-cited violation (NCV)of Technical Specification 5.4.1.a, for the failure to establish adequate maintenance instructions to perform work activities on the division III diesel generator overspeed trip limit switch. Specifically, work orders did not contain adequate instructions to check the overspeed trip switches alignment In accordance with vendor recommendations. As a result, the division III diesel generator was rendered inoperable and unavailable.

Description.

The division III diesel generator is a dual engine, single generator unit in a tandem configuration. Each of the two engines has a mechanical overspeed trip mechanism and an overspeed trip switch. The switches are installed adjacent to each engines overspeed trip lever. In an overspeed event, the mechanical overspeed trip mechanisms actuate. This in turn actuates the respective overspeed trip switch, initiating an electrical trip, and the diesel generator is automatically secured. Actuation of either one of the engines trip switches can successfully initiate an overspeed trip to protect the diesel generator, in the event of a diesel generator overspeed condition. The protective function provided by the overspeed trip remains active in all modes of diesel generator operation.

During the July 13, 2015, monthly surveillance on the division III diesel generator, the diesel inadvertently tripped on overspeed logic. It was determined that the spurious overspeed trip was not caused by an actual overspeed condition. The trip was caused by the overspeed limit switch mechanically disengaging from the overspeed trip lever, which was caused by mechanical wear. Since the overspeed trip lever was no longer in contact with the limit switch, the overspeed logic was fulfilled, and the diesel generator automatically tripped.

The limit switch interfaces with the overspeed lever for both the A and B engines were inspected on December 15, 2013, per Work Orders 307601 and 307598. These work orders were created to address operating experience provided from the diesel generator vendor to the industry in December 2011 per EMD Owners Group Information Bulletin IB11-49. Grand Gulf Nuclear Station captured this operating experience and initiated Condition Report CR-GGN-2011-8269. The inspectors reviewed CR-GGN-2011-8269, and noted that the condition report adequately discussed and evaluated the mechanical wear issues outlined in the operating experience. However, actions taken to develop work orders and provide instructions to inspect for the problem were not adequate to appropriately check and adjust the limit switch.

Analysis.

The failure to establish adequate work instructions to verify the overspeed switch was properly set and adjusted was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment 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, work orders to check the overspeed trip switches alignment did not contain adequate instructions to successfully perform the maintenance. The division III diesel generator was declared inoperable when the diesel spuriously tripped during the monthly surveillance run on July 13, 2015. The inspectors performed the initial significance determination for the division III emergency diesel generator failure. The inspectors used the NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The finding required a detailed risk evaluation because it involved a performance deficiency that represented a loss of the high pressure core spray system following a postulated loss of offsite power because of the failure of the division III diesel generator. The Region IV senior reactor analyst performed a detailed risk evaluation in accordance with NRC Inspection Manual 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The detailed risk evaluation result is a finding of very low safety significance (Green). The calculated change in core damage frequency of 5.0 x 10-7 was dominated by an unrecovered station blackout beyond battery depletion. The analyst determined that the bounding risk of a large, early release of radiation was 9.6 x 10-8. For the details of the analysis, see Attachment 3.

Work orders were developed to address operating experience provided from the diesel generator vendor to the industry in December 2011. The inspectors determined that the cause of the performance deficiency occurred in 2011, and therefore, determined the finding did not have a cross-cutting aspect since it is not indicative of current licensee performance.

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 of Regulatory Guide 1.33, Revision 2. Section 9.a of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for performing maintenance, such that, maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with documented instructions appropriate to the circumstances. The licensee established Work Orders 307598 and 307601 to meet the Regulatory Guide 1.33 requirement.

Contrary to the above, from August 2011 until July 13, 2015, the licensee failed to establish documented instructions appropriate to the circumstances. Specifically, Work Orders 307598 and 307601 failed to ensure operating experience from the diesel generator vendor was incorporated to successfully inspect and setup the overspeed trip mechanism for the division III diesel generator. As a result, the overspeed limit switch disengaged the overspeed lever due to normal wear during the July 13, 2015, monthly surveillance run, and the diesel generator was declared inoperable. Subsequently, the licensee appropriately set the limit switch to overspeed actuating arm engagement, and returned the diesel generator to operable status on July 15, 2015. Because this finding is determined to be of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-GGN-2015-3985, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000416/2015004-4, Failure to Establish Adequate Maintenance Instructions to Perform Work Activities on the Division III Diesel Generator Overspeed Trip Limit Switch.

1R22 Surveillance Testing

Main article: IP 71111.22

a. Inspection Scope

The inspectors observed eight 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 tests:

November 3, 2015, standby service water C pump 1P41-C002, and standby service water loop C return to cooling tower A valve 1P41-F011, quarterly surveillance following washer replacement Reactor coolant system leak detection tests:

December 23, 2015, reactor coolant system leak detection surveillance test Other surveillance tests:

November 13, 2015, settle time testing for control rod 40-41 and monthly operability checks for withdrawal block

November 13, 2015, monthly operability surveillance on control rod 28-21 for withdrawal rod block

November 17, 2015, low pressure coolant injection loop A discharge flow low bypass functional test

November 17, 2015, residual heat removal pump A discharge pressure functional test channel 2A

December 11, 2015, division III 125-volt DC battery inter-cell connection resistance surveillance tests on December 6, 2013, and February 25, 2015

December 23, 2015, reactor coolant dose equivalent iodine sample surveillance test The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.

These activities constituted completion of eight surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.

b. Findings

No findings were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation

Main article: IP 71114.06

Training Evolution Observation

a. Inspection Scope

On November 16, 2015, the inspectors observed simulator-based licensed operator requalification training that included implementation of the licensees emergency plan.

The inspectors verified that the licensees 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 evaluators and entered into the corrective action program for resolution.

These activities constituted completion of one training observation sample, as defined in Inspection Procedure 71114.06.

b. Findings

No findings were identified.

1EP7 Exercise Evaluation - Hostile Action Event

Main article: IP 71114.07

a. Inspection Scope

The inspectors observed the October 21, 2015, biennial emergency plan exercise to verify the exercise acceptably tested the major elements of the emergency plan, provided opportunities for the emergency response organization to demonstrate key skills and functions, and demonstrated the licensees ability to coordinate with offsite emergency responders. The scenario simulated:

A large aircraft threat to the site

An impact of the aircraft to the site protected area

Damage to the fire water pump house and tanks and demineralized water tank

A loss of the circulating water and condensate systems due to debris from the aircraft impacting non-safety electrical buses

Injured and deceased plant employees The exercise scenario was developed to demonstrate the licensees capability to implement its emergency plan under conditions of uncertain physical security.

During the exercise the inspectors observed activities in the control room simulator and the following emergency response facilities:

Alternate Technical Support Center

Alternate Operations Support Center

Backup Emergency Operations Facility

Central and/or Secondary Alarm Station

Incident Command Post The inspectors focused their evaluation of the licensees performance on event classification, offsite notification, recognition of offsite dose consequences, development of protective action recommendations, staffing of alternate emergency response facilities, and the coordination between the licensee and offsite agencies to ensure reactor safety under conditions of uncertain physical security.

The inspectors also assessed recognition of, and response to, abnormal and emergency plant conditions, the transfer of decision-making authority and emergency function responsibilities between facilities, onsite and offsite communications, protection of plant employees and emergency workers in an uncertain physical security environment, emergency repair evaluation and capability, and the overall implementation of the emergency plan to protect public health, safety, and the environment. The inspectors reviewed the current revision of the facility emergency plan, emergency plan implementing procedures associated with operation of the licensees primary and alternate emergency response facilities, and procedures for the performance of associated emergency and security functions.

The inspectors attended the post-exercise critiques in each emergency response facility to evaluate the initial licensee self-assessment of exercise performance. The inspectors also attended a subsequent formal presentation of critique items to plant management.

The specific documents reviewed during this inspection are listed in the attachment.

The inspectors reviewed the scenario of previous biennial exercises and licensee drills conducted between September 2013 and October 2015 to determine whether the October 21, 2015, exercise was independent and avoided participant preconditioning in accordance with the requirements of 10 CFR Part 50, Appendix E, IV.F(2)(g). The inspectors also compared observed exercise performance with corrective action program entries and after-action reports for drills and exercises conducted between September 2013 and October 2015 to determine whether identified weaknesses had been corrected in accordance with the requirements of 10 CFR 50.47(b)(14), and 10 CFR Part 50, Appendix E, IV.F.

These activities constituted completion of one exercise evaluation sample, as defined in Inspection Procedure 71114.07.

b. Findings

No findings were identified.

1EP8 Exercise Evaluation - Scenario Review

Main article: IP 71114.08

a. Inspection Scope

The licensee submitted the preliminary exercise scenario for the October 21, 2015, biennial exercise to the NRC on August 14, 2015, in accordance with the requirements of 10 CFR Part 50, Appendix E, IV.F(2)(b). The inspectors performed an in-office review of the proposed scenario to determine whether it would acceptably test the major elements of the licensees emergency plan, and provide opportunities for the emergency response organization to demonstrate key skills and functions.

These activities constituted completion of one exercise evaluation - scenario review sample, as defined in Inspection Procedure 71114.08.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

Main article: IP 71124.01

a. Inspection Scope

The inspectors assessed the licensees performance in assessing the radiological hazards in the workplace associated with licensed activities. The inspectors assessed the licensees implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors walked down various portions of the plant and performed independent radiation dose rate measurements. The inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors reviewed licensee performance in the following areas:

The hazard assessment program, including a review of the licensees evaluations of changes in plant operations and radiological surveys to detect dose rates, airborne radioactivity, and surface contamination levels

Instructions and notices to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions

Programs and processes for control of sealed sources and release of potentially contaminated material from the radiologically controlled area, including survey performance, instrument sensitivity, release criteria, procedural guidance, and sealed source accountability

Radiological hazards control and work coverage, including the adequacy of surveys, radiation protection job coverage and contamination controls, the use of electronic dosimeters in high noise areas, dosimetry placement, airborne radioactivity monitoring, controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools, and posting and physical controls for high radiation areas and very high radiation areas

Radiation worker and radiation protection technician performance with respect to radiation protection work requirements

Audits, self-assessments, and corrective action documents related to radiological hazard assessment and exposure controls since the last inspection These activities constituted completion of one sample of radiological hazard assessment and exposure controls, as defined in Inspection Procedure 71124.01.

b. Findings

No findings were identified.

2RS2 Occupational ALARA Planning and Controls

Main article: IP 71124.02

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining occupational individual and collective radiation exposures as low as is reasonably achievable (ALARA). During the inspection, the inspectors interviewed licensee personnel and reviewed licensee performance in the following areas:

Site-specific ALARA procedures and collective exposure history, including the current 3-year rolling average, site-specific trends in collective exposures, and source-term measurements

ALARA work activity evaluations/postjob reviews, exposure estimates, and exposure mitigation requirements

The methodology for estimating work activity exposures, the intended dose outcome, the accuracy of dose rate and man-hour estimates, and intended versus actual work activity doses and the reasons for any inconsistencies

Records detailing the historical trends and current status of tracked plant source terms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry

Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas

Audits, self-assessments, and corrective action documents related to ALARA planning and controls since the last inspection These activities constituted completion of one sample of occupational ALARA planning and controls, as defined in Inspection Procedure 71124.02.

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

4OA1 Performance Indicator Verification

Main article: IP 71151

.1 Reactor Coolant System Specific Activity (BI01)

.7 Radiological Effluent Technical Specifications (RETS)/Offsite Dose Calculation Manual

(ODCM) Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors reviewed corrective action program records for liquid or gaseous effluent releases that occurred between January 1, 2014, and September 30, 2015, and were reported to the NRC to verify the performance indicator data. 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 reported data.

These activities constituted verification of the radiological effluent technical specifications (RETS)/offsite dose calculation manual (ODCM) radiological effluent occurrences performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

Main article: IP 71152

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensees corrective action program and periodically attended the licensees 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 licensees problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Semiannual Trend Review

a. Inspection Scope

On December 22, 2015, the inspectors completed their review of the licensees corrective action program, performance indicators, system health reports, backlog trending reports, and other documentation to identify trends that might indicate the existence of a more significant safety issue. The inspectors verified that the licensee was taking corrective actions to address identified adverse trends.

These activities constituted completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.

b. Observations and Assessments On December 21, 2015, the inspectors completed a review of an adverse trend in preventative maintenance activities that are past the original due date and into the grace period. This trend was documented in Condition Reports CR-GGN-2015-06388 and CR-GGN-2015-07276. The inspectors also reviewed several condition reports that identified late preventative maintenance activities, the licensees backlog trend reports and the current preventative maintenance schedule for deep grace activities.

Of the Entergy Fleet, Grand Gulf Nuclear Station currently has the highest number of preventative maintenance activities in the grace period. The licensee currently has 212 preventative maintenance activities in the grace period, and the activities span the disciplines of electrical, mechanical, and instrumentation and controls, with 128 activities being in electrical.

Of the 212 activities, 14 preventative maintenance activities are greater than 50 percent over the allotted grace period. The licensee currently has all of these preventative maintenance activities scheduled with a clear plan to complete prior to the end of the allotted grace period. If the licensee does not complete a preventative maintenance activity within the grace period, a corrective action is assigned to engineering to evaluate the impact to the equipment and ultimately decide if the equipment needs to be taken out of service prior to the next scheduled system outage.

Corrective actions associated with the adverse trend are:

The licensee has taken action already to increase staff allocation in the electrical field so that there are more staff to accomplish the preventative maintenance tasks. Currently, the licensee has identified that there is a shortage of electrical workers and is actively working to increase staff.

The licensee will be changing their work scheduling process to go from a 12-week schedule to a 13-week schedule. This allows for an extra week to finish any activities that could not be completed during the appropriate maintenance windows. They are also focusing on ensuring that all required preventative maintenance activities are scheduled during the system outage.

The licensee changed their look ahead schedule from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This allows the staff to understand impacts of deferring work since they will now understand and know their future work load.

The licensee is changing the work week managers work schedule from Monday to Monday to a Friday to Friday week. Previously, items were being deferred and that communication was not always clear to the oncoming work week manager following the weekend. This change will ensure that the work week manager is completing all assigned work tasks prior to leaving site for the weekend, resulting in fewer turnover issues.

c. Findings

No findings were identified.

4OA5

OTHER ACTIVITIES

Operation of an Independent Spent Fuel Storage Facility Installation (ISFSI) at Operating Plants (60855.1)a. Operation of an ISFSI Inspection Scope A routine ISFSI inspection was conducted of the Grand Gulf Nuclear Station ISFSI on October 26-29, 2015, by a Region IV Division of Nuclear Material Safety inspector. The inspector observed loading operations and reviewed selected licensee loading, processing, and heavy load procedures associated with the licensees current dry fuel storage loading campaign. The inspector performed a review of the fuel assemblies selected for placement into dry fuel storage for the current ISFSI campaign to verify that the licensee was loading fuel in accordance with the Holtec Certificate of Compliance (CoC) 1014 Approved Contents. The inspector reviewed documents including: 1) the multi-purpose canister (MPC) loading maps; and 2) the fuel assembly qualification information from the approved contents consisting of the assembly decay heat (kW),cooling time (years), average U-235 enrichment (percent), and cumulative burnup (MWd/MTU).

Various loading activities were observed by the NRC inspector during the course of the routine ISFSI inspection. The licensee was in the process of loading canister #25 at the time of the inspection. There was a previously loaded HI-STORM that remained inside the fuel building during the loading operations due to weather related delays that prevented the previously loaded HI-STORM from being placed onto the ISFSI pad.

Selected welding and non-destructive examination evolutions were observed during the loading activities associated with canister #25. An automated welding process was used to perform the MPC lid to shell closure weld. The welding machine utilized a single weld head for the lid-to-shell weld. The welders operated their equipment remotely in an area of low radiological dose while monitoring the progress and performance of the machine using dual video monitors. Hydrogen monitoring was performed during the welding of the root pass. In addition, the NRC inspector observed the non-destructive liquid dye penetrant exams conducted on the first pass of the lid-to-shell weld and on the final lid closure weld after the hydrostatic pressure testing of the weld had been completed. All of the non-destructive examinations (NDE) observed by the NRC inspector passed with clear results. All noted indications (both relevant and non-relevant) were included in the documentation of the weld traveler as a permanent record by the NDE technician.

The NRC inspector observed the haul path hazard identification walk-down by the ISFSI manager and observed the vertical cask transporter lift and transport the previously loaded HI-STORM 100 cask (with canister #24) out of the fuel building and along the designated haul path to the ISFSI pad.

The NRC inspector verified the radiological conditions at the ISFSI through a review of the most recent radiological survey and a walk-down of the ISFSI pad with radiation survey instrumentation. The NRC inspector accompanied a licensing representative and one radiation protection (RP) technician during the walk-down of the ISFSI pad. The ISFSI pad was clear of vegetative overgrowth and there were no unexpected combustible or flammable items present on the storage pad. The ISFSI pad contained 23 HI-STORM-100 casks which were in good physical condition. ISFSI boundary radiological measurements were taken by the RP technician with a Geiger-Mueller detector to record gamma exposure rates. The NRC inspector carried a Ludlum Model 19 sodium-iodide survey meter (NRC #033906, calibration due 3/13/2016) and recorded measurements at the ISFSI boundary. The measurements taken by the NRC inspector and RP technician confirmed the measurements recorded on the most recent ISFSI site survey. The radiological conditions in and around the ISFSI were as expected for the age and heat-load of the 23 loaded spent fuel storage casks. Annual Radiological Environmental Operating Reports were reviewed for the last two years. The reports documented the dose equivalent to any real individual located beyond the site controlled area had been well below the 10 CFR 72.104(a)(2) requirement of less than 25 mrem per year.

The NRC inspector reviewed a randomly selected month of HI-STORM 100 vent surveillance records to ensure that the Holtec CoC 1014 Technical Specifications (TS)requirements were being met for fuel stored on the ISFSI pad.

The inspector requested documentation related to maintenance, modifications, and safety evaluations performed for the fuel building cask handling crane. Documents were provided that demonstrated the fuel building cask handling crane was inspected on an annual basis in accordance with the safety standards of the American Society of Mechanical Engineers (ASME) B30.2, "Overhead and Gantry Cranes," prior to the 2015 loading campaign.

A list of Condition Reports (CRs) issued since the last NRC inspection conducted in June 2014, was provided by the licensee for the cask handling crane and the ISFSI operations. When a problem was identified the licensee would document the issue as a CR for placement in the licensees corrective action program. Of the list of CRs provided relating to the ISFSI and the cask handling cranes, eleven were selected by the NRC inspector for further review. The CRs were well documented and properly categorized based on the safety significance of the issue. The corrective actions taken were appropriate to the situations. Based on the level of detail of the corrective action reports, the licensee demonstrated good attention to detail in regard to the maintenance and operation of their ISFSI program and the cask handling crane. No NRC safety concerns were identified related to the CRs selected during this inspection.

The licensees 10 CFR 72.48 screenings and evaluations for ISFSI program changes since the last NRC routine ISFSI inspection were reviewed to determine compliance with regulatory requirements. The licensee had performed one 72.48 screen and no full 72.48 evaluations since the last NRC inspection. The licensee had not performed any 10 CFR 50.59 screenings or evaluations for the fuel building cask handling crane since the last inspection. The 72.48 screening that was reviewed was determined to have been adequately evaluated by the licensee.

An on-site review of the quality assurance audit and quality assurance surveillance reports related to dry cask storage activities at the ISFSI was performed by the NRC inspector. The QA audit report resulted in several condition reports. The NRC inspector reviewed the corrective actions resulting from the condition reports to ensure that the identified deficiencies were properly categorized based on their safety significance and properly resolved. All audit identified deficiencies had been properly categorized and resolved by the licensee. The licensee had not performed any QA surveillances since the last inspection.

The inspector observed that the licensee had met the licensing requirements for the documents and activities reviewed associated with the dry cask storage activities at Grand Gulf Nuclear Station.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On September 8, 2015, the inspectors discussed the in-office review of the preliminary scenario for the 2015 biennial exercise, submitted August 14, 2015, with Mr. D. Ellis, Manager, Emergency Planning, 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.

On October 23, 2015, the inspectors presented the results of the onsite inspection of the biennial emergency preparedness exercise conducted October 21, 2015, to Mr. K. Mulligan, Site Vice President, 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.

On October 29, 2015, the inspector presented the results of the ISFSI inspection to Mr. K. Mulligan, Site Vice President, 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.

On November 19, 2015, the inspectors presented the radiation safety inspection results to Mr. V. Fallacara, General Manager of Plant Operations, 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.

On January 7, 2015, the inspectors presented the inspection results to Mr. K. Mulligan, Site Vice President, and other members of the licensee staff, and on February 11, 2016, the inspectors presented the final inspection results to Mr. T. Coutu, Director, Regulatory and Performance Improvement. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

4OA7 Licensee-Identified Violations

The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as non-cited violations.

1. Technical Specification 3.6.1.3, Surveillance Requirement 3.6.1.3.9, requires, verification of combined leakage rate of 1 gallon per minute times the total number of primary containment isolation valves through hydrostatically tested lines that penetrate the primary containment is not exceeded when these isolation valves are tested at greater than or equal to 1.1 times peak containment pressure. Contrary to the above, since June 6, 2012, the licensee failed to verify combined leakage rate of 1 gallon per minute times the total number of primary containment isolation valves through hydrostatically tested lines that penetrate the primary containment is not exceeded when these isolation valves are tested at greater than or equal to 1.1 times peak containment pressure. Specifically, the post-extended power uprate peak containment pressure analyzed increased to 14.8 psig, resulting in a new required test pressure of 16.28 psig.

The licensee did not test primary containment isolation valves 1P11F130 and 1P11F131 using the new higher pressure. The licensee subsequently declared the valves inoperable and tested the two valves using the new peak containment pressure. The valves passed the surveillance test and were declared operable at 11:01 am on October 15, 2015. This finding was entered in the licensee's corrective action program as Condition Report CR-GGN-2015-05072.

The finding is more than minor because it was associated with the barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment)protect the public from radionuclide releases caused by accidents or events.

Specifically, the licensee never performed Technical Specification Surveillance Requirement 3.6.1.3.9, and therefore did not have presumption of operability to provide the reasonable assurance that containment would protect the public from radionuclide releases caused by accidents or events. The significance of the finding was assessed using Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, and it was determined to be of very low safety significance (Green).

2. Title 10 CFR 50.54(hh)(1)(iv) and

(vi) require, in part, that licensees implement onsite actions necessary to enhance the capability of the facility to mitigate the consequences of an aircraft impact; and procedures for dispersal of equipment and personnel.

Regulatory Guide 1.214, "Response Strategies for Potential Aircraft Threats,"

Section 7.1, states that to meet the dispersal requirement, licensee should include security personnel for accomplishing post-impact meditative actions in aircraft threat procedures. It further states, to include suitable locations to which those resources can be repositioned to increase survivability. Contrary to the above, on October 21, 2015, during the Grand Gulf Nuclear Station's biennial NRC evaluated exercise, the licensee failed to implement onsite actions necessary to enhance the capability of the facility to mitigate the consequences of an aircraft impact and did not have procedures for the dispersal of equipment and personnel. Specifically, during an emergency preparedness exercise observed by NRC, the licensee had not established an adequate process to use upon receiving potential aircraft threat warnings (simulated) from the NRC, to decide when or if to disperse or reposition security personnel to increase survivability. This finding was entered in the licensee's corrective action program as Condition Report CR-GGN-2015-06195.

The finding is more than minor because if left uncorrected, it would have the potential to lead to a more significant security or safety concern; this failure could potentially and adversely affect survivability of security response personnel in the flight path of a potential aircraft threat as well as the capability to take appropriate actions to ensure adequate security resources when mitigating the consequences of an aircraft impact.

The significance of the finding was assessed using NRC IMC 0609, Appendix E, Part I, Baseline Security Significance Determination Process, and it was determined to be of very low security significance (Green).

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

R. Benson, Superintendent, Radiation Protection

C. Boschetti, Manager, Nuclear Oversight

K. Boudreaux, Manager, System Engineering

A. Burks, Supervisor, Radiation Protection

D. Burnett, Director, Emergency Response, Corporate Operations Support

R. Busick, Manager, Operations

T. Coles, Engineer 1, Regulatory Assurance

T. Coutu, Director, Regulatory and Performance Improvement

C. Dawson II, Emergency Planner, Emergency Planning

J. Dorsey, Manager, Security

D. Ellis, Senior Emergency Planner/Acting Manager Emergency Planning

V. Fallacara, General Manager Plant Operations

E. Garrison, Acting Manager, Training

M. Goodwin, Manager, Operations Support

B. Grant, Manager, Production

G. Hawkins, Manager, Site Projects/Maintenance Services

M. Lanni, Supervisor, Radiation Protection

M. Larson, Supervisor, Radiation Protection

C. Lewis, Manager, Operational IT

R. Meister, Senior Specialist, Regulatory Assurance

R. Miller, Manager, Radiation Protection

R. Millison, Coordinator, Site Vice President

M. Milly, Senior Manager, Maintenance

T. Moncure, Senior Technician, Radiation Protection

K. Mulligan, Site Vice President

J. Nadeau, Manager, Regulatory Assurance

E. Riggs, Manager, ISFSI

P. Salgado, Manager, Performance Improvement

R. Scarbrough, NRC Interface

J. Seiter, Manager, Emergency Planning

P. Stokes, Supervisor, Radiation Protection

R. Sumrall, Manager, Chemistry

S. Sweet, Engineer, Licensing

R. Vandenakker, Emergency Planner, Emergency Planning

M. Vanslyke, Fleet Supervisor, Entergy Dry Cask Storage

B. Wertz, Outage Manager, Production

P. Williams, Director, Engineering

R. Young, Auditor, Nuclear Oversight

NRC Personnel

C. Stott, Reactor Inspector

T. Sullivan, Operations Inspector

D. Loveless, Senior Reactor Analyst

Attachment 1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

Failure to Have Appropriate Instructions for Preventative Maintenance on the Division I Diesel Generator Simulated Run

05000416/2015004-01 NCV (Section 1R12)

Failure to Timely Enter Technical Specification Surveillance

05000416/2015004-02 NCV Requirement 3.0.1 (Section 1R13)

Failure to Make a Required Eight-Hour Report for Loss of

05000416/2015004-03 NCV Safety Function (Section 1R15)

Failure to Establish Adequate Maintenance Instructions to

05000416/2015004-04 NCV Perform Work Activities on the Division III Diesel Generator Overspeed Trip Limit Switch (Section 1R19)

LIST OF DOCUMENTS REVIEWED