IR 05000416/2006002
Download: ML061290037
Text
May 8, 2006
George A. Williams, Site Vice PresidentGrand Gulf Nuclear Station Entergy Operations, Inc.
P.O. Box 756 Port Gibson, MS 39150
SUBJECT: GRAND GULF NUCLEAR STATION - NRC INTEGRATED INSPECTIONREPORT 05000416/2006002
Dear Mr. Williams:
On March 31, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat your Grand Gulf Nuclear Station facility. The enclosed integrated report documents theinspection findings, which were discussed on April 11, 2006, with you and other members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.This report documents four NRC identified and self-revealing findings of very low safetysignificance (Green). Three of these findings were determined to involve violations of NRC requirements; however, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating these findings as noncitedviolations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. If you contest 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, U.S. Nuclear Regulatory Commission Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the Grand Gulf NuclearStation facility.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be made available electronically for public inspection Entergy Operations, Inc.- 2 -in the NRC Public Document Room or from the Publicly Available Records (PARS) componentof NRC's document 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/Kriss M. Kennedy, ChiefProject Branch C Division of Reactor ProjectsDocket: 50-416License: NPF-29
Enclosure:
Inspection Report 05000416/2006002
w/Attachment:
Supplemental Informationcc w/enclosure:Senior Vice President and Chief Operating Officer Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995Wise, Carter, Child & CarawayP.O. Box 651 Jackson, MS 39205Winston & Strawn LLP1700 K Street, N.W.
Washington, DC 20006-3817Jay Barkley, ChiefEnergy & Transportation Branch Environmental Compliance and Enforcement Division Mississippi Department of Environmental Quality P.O. Box 10385 Jackson, MS 39289-0385President, District 1Claiborne County Board of Supervisors P.O. Box 339 Port Gibson, MS 39150 Entergy Operations, Inc.- 3 -General ManagerGrand Gulf Nuclear Station Entergy Operations, Inc.
P.O. Box 756 Port Gibson, MS 39150The Honorable Charles C. Foti, Jr.Attorney General Department of Justice State of Louisiana P.O. Box 94005 Baton Rouge, LA 70804-9005 Governor Haley BarbourOffice of the Governor State of Mississippi P.O. Box 139 Jackson, MS 39205Jim Hood, Attorney GeneralState of Mississippi P.O. Box 220 Jackson, MS 39225 Dr. Brian W. AmyState Health Officer State Board of Health P.O. Box 1700 Jackson, MS 39215 Robert W. Goff, Program DirectorDivision of Radiological Health Mississippi Dept. of Health P.O. Box 1700 Jackson, MS 39215-1700DirectorNuclear Safety & Licensing Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213-8298Director, Nuclear Safety and Regulatory Affairs Entergy Operations, Inc.
P.O. Box 756 Port Gibson, MS 39150 Entergy Operations, Inc.- 4 -ChairpersonDenton Field Office Chemical and Nuclear Preparedness and Protection Division Office of Infrastructure Protection Preparedness Directorate Dept. of Homeland Security 800 North Loop 288 Federal Regional Center Denton, TX 76201-3698Radiological Assistance Committee ChairChemical and Nuclear Preparedness and Protection Division Atlanta Field Office Dept. of Homeland Security 3003 Chamblee-Tucker Road Atlanta, GA 30341 Entergy Operations, Inc.- 5 -Electronic distribution by RIV:Regional Administrator (BSM1)DRP Director (ATH)DRS Director (DDC)DRS Deputy Director (RJC1)Senior Resident Inspector (GBM)Branch Chief, DRP/C (KMK)Senior Project Engineer, DRP/C (WCW)Team Leader, DRP/TSS (RLN1)RITS Coordinator (KEG)DRS STA (DAP)S. O'Connor, OEDO RIV Coordinator (SCO)ROPreports GG Site Secretary (NAS2)W. A. Maier, RSLO (WAM)SUNSI Review Completed: kmk___ADAMS: YesG No Initials: __kmk__ Publicly Available G Non-Publicly Available G Sensitive Non-SensitiveR:\_REACTORS\GG\2006\GG2006-02RP-GBM.wpdRIV:RI:DRP/CSRI:DRP/CC:SPE:DRP/CC:DRS/EB1C:DRS/PSBAJBarrett GBMillerWCWalkerJAClarkMPShannon E - KMKennedy E - KMKennedy /RA/ ATGody for LCCarson for5/8/065/8/065/3/065/5/065/5/06C:DRS/OBC:DRS/EB2C:DRP/CATGodyLJSmithKMKennedy /RA/ /RA/ /RA/5/5/065/5/065/8/06OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure-1-U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket:50-416Licenses:NPF-29 Report No.:05000416/2006002 Licensee:Entergy Operations, Inc.
Facility:Grand Gulf Nuclear StationLocation:Waterloo Road Port Gibson, Mississippi 39150Dates:January 1 through March 31, 2006 Inspectors:G. Miller, Senior Resident InspectorA. Barrett, Resident Inspector M. Sitek, Resident Inspector L. Carson, Senior Health Physics Inspector N. O'Keefe, Senior Reactor InspectorApproved By:Kriss M. Kennedy, ChiefProject Branch C Division of Reactor Projects Enclosure-2-
SUMMARY OF FINDINGS
....................................................1 REACTOR SAFETY........................................................51R01 Adverse Weather Protection........................................51R04 Equipment Alignment.............................................61R05 Fire Protection..................................................71R06 Flood Protection Measures.........................................81R11 Licensed Operator Requalification...................................91R12 Maintenance Effectiveness.........................................91R13 Maintenance Risk Assessments and Emergent Work Control..............91R14 Personnel Performance During Nonroutine Plant Evolutions..............131R15 Operability Evaluations...........................................131R19 Postmaintenance Testing.........................................141R22 Surveillance Testing.............................................151R23 Temporary Plant Modification......................................161EP6 Drill Evaluation.................................................16RADIATION SAFETY.......................................................172OS1 Access Control to Radiologically Significant Areas.....................172OS2 ALARA Planning and Controls.....................................18OTHER ACTIVITIES........................................................194OA1 Performance Indicator Verification..................................194OA2 Identification and Resolution of Problems............................204OA5 Other Activities.................................................23 4OA6 Meetings, Including Exit..........................................25ATTACHMENT: SUPPLEMENTAL INFORMATION...............................A-1KEY POINTS OF CONTACT................................................A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED...........................A-1LIST OF DOCUMENTS REVIEWED..........................................A-2 LIST OF ACRONYMS......................................................A-5 Enclosure-3-SUMMARY OF FINDINGSIR 05000416/2006002; 01/01/06 - 03/31/06; Grand Gulf Nuclear Station -- Integrated Residentand Regional Report; Maintenance Effectiveness, Identification and Resolution of Problems,
Other Activities.This report covered a 3-month period of inspection by resident inspectors and Regional officeinspectors. The inspection identified four Green findings, three of which were noncited violations. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management's review. The NRC's program for overseeing the safeoperation of commercial nuclear power reactors is described in NUREG-1649, "ReactorOversight Process," Revision 3, dated July 2000.A.NRC-Identified and Self-Revealing Findings
Cornerstone: Initiating Events
- Green.
The inspectors identified two examples of a noncited violation of 10 CFR 50.65,"Maintenance Rule," for failing to include maintenance that could increase the likelihoodof an initiating event in the plant risk assessment. On February 2, 2006, and again on March 28, 2006, the licensee's risk assessment did not include maintenance activities that increased the likelihood of a reactor scram. The licensee entered this into their corrective action program as Condition Reports CR-GGN-2006-1041 and CR-GGN-2006-1277.This finding is more than minor since the maintenance that was performed increased thelikelihood of an initiating event. Using Inspection Manual Chapter 0609 Appendix K,
"Maintenance Risk Assessment and Risk Management Significance Determination Process," the finding is of very low safety significance since in both cases the change in incremental core damage probability and incremental large early release probabilitywere less than 1E-6 and 1E-7, respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnelerror (Section 1R13.1).*Green. The inspectors reviewed a self-revealing finding for a failure to follow aprocedure that resulted in a significant plant service water header leak. The licensee failed to perform an adequate review of documents to identify potential hazards as required by Procedure EN-S-112, "Trenching, Excavation and Ground Penetrating Activities," Revision 2. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-0219.This finding is more than minor since it was associated with the human performanceattribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of aSignificance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant Enclosure-4-accident, did not contribute to a loss of mitigation equipment, and did not increase thelikelihood of a fire or internal/external flood. The cause of this finding has human performance crosscutting aspects associated with a failure to follow procedures(Section 1R13.2).
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a Green noncited violation of 10 CFR Part 50,Appendix B, Criterion XVI, for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation. The licensee identified the design deficiency on April 30, 1999, and issued compensatory actions for the operators to manually transfer high pressure core spray and reactor core isolation cooling from the condensate storage tank to the suppression pool in the event of failureof the tank. The licensee corrected the design deficiency on December 8, 2005. The licensee entered this issue in their corrective action program as CR-GGN-2006-1096. This finding is more than minor because it affected the design control attribute of themitigating systems cornerstone and affected the cornerstone objective to ensure theavailability of systems that respond to initiating events. The finding was of very lowsafety significance because it was a design deficiency that did not result in a loss of operability. This finding had crosscutting aspects associated with problem identificationand resolution in that station personnel did not implement corrective actions in a timely manner (Section 4OA2).
- Green.
The inspectors identified a Green noncited violation for failure to have analternative shutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. The licensee entered this into their corrective action program as CR-GGN-2005-1854.This finding is more than minor because it affected the mitigating systems cornerstoneobjective for the procedure quality and protection from external factors attributes. A Region IV Senior Reactor Analyst made a visit to the site during the week of January 30, 2006. Through discussions with engineers and walkdowns in the plant, the Senior Reactor Analyst determined that there is a credible fire scenario which could simultaneously cause a control room evacuation, a loss of offsite power, and prevent automatic starting and loading of the Division 1 emergency diesel generator. This issue was categorized as a postfire safe shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were adequate toovercome the procedure deficiency. Therefore, this issue screened as having very low safety significance in Phase 1 of Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process" (Section 4OA3.1).
B.Licensee-Identified Violations
None.
Enclosure-5-
REPORT DETAILS
Summary of Plant StatusGrand Gulf Nuclear Station started the inspection period at approximately 87 percent powerdue to a planned control rod pattern adjustment and power suppression testing for a suspected fuel leak. The reactor returned to full power on January 2, 2006. On March 22, 2006, power was reduced to approximately 50 percent due to a reactor feed pump trip. The plant returned to full power on March 26, 2006. Over the balance of the inspection period, the plant remainedat or near full power except for planned control rod pattern adjustments and control rod drive maintenance and testing.1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, Barrier Integrity1R01Adverse Weather Protection (71111.01).1Readiness For Seasonal Susceptibilities
a. Inspection Scope
The inspectors completed a review of the licensee's readiness for seasonalsusceptibilities involving extremely low temperatures. The inspectors: (1) reviewed plant procedures, the Updated Final Safety Analysis Report (UFSAR), and Technical Specifications (TS) to ensure that operator actions defined in adverse weather procedures maintained the readiness of essential systems; (2) walked down portions ofthe three systems listed to ensure that adverse weather protection features were sufficient to support operability, including the ability to perform safe shutdown functions;(3) evaluated operator staffing levels to ensure the licensee could maintain the readiness of essential systems required by plant procedures; and (4) reviewed thecorrective action program (CAP) to determine if the licensee identified and corrected problems related to adverse weather conditions. January 6, 2006, plant service water systemJanuary 9, 2006, emergency diesel generators (EDGs)January 12, 2006, standby service water systemDocuments reviewed by the inspectors are listed in the attachment. The inspectorscompleted three samples.
b. Findings
Introduction.
An unresolved item was identified for inadequate design control of freezeprotection equipment in the diesel generator building corridor.Description. Prior to the onset of freezing conditions, the licensee installs temporaryheaters in the diesel generator breezeway, an enclosed space between the diesel
-6-building and the auxiliary building with open grating at each end. Fire protection pipingand safety-related standby service water piping in the breezeway are also provided with heat tracing for cold weather protection.During a walkdown of the breezeway on January 11, 2006, the inspectors questionedwhether the temporary heaters installed were of sufficient size to protect piping in the breezeway from freezing. The licensee could not produce any calculations or testing documentation to justify the sizing of the heaters, stating instead that the heaters werenot needed to protect safety-related equipment, so the size was selected based on engineering judgement.The inspectors found that a corrective action to Condition Report CR-GGN-2002-2250identified reliability issues with the heat tracing installed on safety-related piping andconcluded that the best means for ensuring freeze protection for all piping in the breezeway was through the use of area heating via space heaters. The inspectors concluded that the corrective actions of condition report CR-GGN-2002-2250 effectively abandoned the heat tracing in place and instead credited the space heaters as supplying freeze protection for the safety-related piping in the breezeway.The licensee initiated Condition Report CR-GGN-2006-1518 to evaluate the currentcondition of the heat tracing in the diesel generator building breezeway and to assess the sizing of the area heaters.Analysis. The failure to verify the adequacy of the design of the area heaters installedto replace the heat trace on safety-related piping is a performance deficiency. This finding is associated with the design control attribute of the Mitigating Systemscornerstone and is more than minor since it affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiatingevents. This issue remains unresolved pending review of the licensee's evaluation associated with Condition Report CR-GGN-2006-1518.Enforcement. This finding is unresolved pending further review of the licensee'sevaluation associated with condition report CR-GGN-2006-1518. URI 05000416/2006002-01, Inadequate Design Control for Freeze Protection in the DieselBuilding Breezeway.
1R04 Equipment Alignment (71111.04).1Partial System Walkdowns
a. Inspection Scope
The inspectors: (1) walked down portions of the three listed risk important systems andreviewed plant procedures and documents to verify that critical portions of the selected systems were correctly aligned; and (2) compared deficiencies identified during thewalkdown to the licensee's UFSAR and CAP to ensure problems were being identified and corrected.
-7-January 12, 2006, the inspectors walked down Train A of the control room airconditioning system while Train B was out of service for planned maintenance.February 15, 2006, the inspectors walked down Train B of the standby gastreatment system while Train A was out of service for planned maintenance.March 9, 2006, the inspectors walked down the Division I EDG while theDivision II EDG was out of service for planned maintenance.Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted three samples of risk significant systems.
b. Findings
No findings of significance were identified..2Complete System Walkdown
a. Inspection Scope
The inspectors: (1) reviewed plant procedures, drawings, the UFSAR, TSs, and vendormanuals to determine the correct alignment of the high pressure core spray system;(2) reviewed outstanding design issues, operator workarounds, and UFSAR documents to determine if open issues affected the functionality of the high pressure core spray system; and (3) verified that the licensee was identifying and resolving equipmentalignment problems. Documents reviewed by the inspectors included:M-1086, P&I Diagram High Pressure Core Spray Unit 1, Revision 30 04-1-01-E22-1, High Pressure Core Spray System, Revision 108 The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
a. Inspection Scope
Quarterly InspectionThe inspectors walked down the six listed plant areas to assess the material condition ofactive and passive fire protection features and their operational lineup and readiness.
The inspectors: (1) verified that transient combustibles and hot work activities were controlled in accordance with plant procedures; (2) observed the condition of fire detection devices to verify they remained functional; (3) observed fire suppression systems to verify they remained functional and that access to manual actuators was
-8-unobstructed; (4) verified that fire extinguishers and hose stations were provided at theirdesignated locations and that they were in a satisfactory condition; (5) verified that passive fire protection features (electrical raceway barriers, fire doors, fire dampers, steel fire proofing, penetration seals, and oil collection systems) were in a satisfactorymaterial condition; (6) verified that adequate compensatory measures were established for degraded or inoperable fire protection features and that the compensatory measureswere commensurate with the significance of the deficiency; and (7) reviewed the UFSAR to determine if the licensee identified and corrected fire protection problems. Division I EDG room (Room 1D302)Division II EDG room (Room1D303)Division III EDG room (Room 1D304)EDG building fresh air corridor (Room 1D301)Control room air conditioning and fresh air system Train A room (Room OC302)Control room air conditioning and fresh air system Train B room (Room OC303)Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted six samples.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures (71111.06).1Semiannual Internal Flooding
a. Inspection Scope
The inspectors: (1) reviewed the UFSAR, the flooding analysis, and plant procedures toassess seasonal susceptibilities involving internal flooding; (2) reviewed the UFSAR andCAP to determine if the licensee identified and corrected flooding problems; (3) verified that operator actions for coping with flooding can reasonably achieve the desired outcomes; and (4) walked down the two below listed areas to verify the adequacy of:
(a) equipment seals located below the floodline, (b) floor and wall penetration seals, (c) watertight door seals, (d) common drain lines and sumps, (e) sump pumps, level alarms, and control circuits, and (f) temporary or removable flood barriers. March 14, 2006, Reactor heat removal Train C pump room (1A118) and pipingroom (1A116) Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted one sample.
b. Findings
No findings of significance were identified.
-9-1R11Licensed Operator Requalification (71111.11)
a. Inspection Scope
The inspectors observed testing and training of senior reactor operators and reactoroperators to assess training, operator performance, and the evaluator's critique. Thetraining scenario, GSMS-LOR-HIT04, Revision 0, involved an anticipated transient without scram (ATWS) with inadvertent high pressure core spray initiation, diesel trip, scram discharge volume leak and subsequent containment pressure increase, and initiation of containment spray. The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors reviewed the following two maintenance activities in order to: (1) verifythe appropriate handling of structure, system, and component (SSC) performance orcondition problems; (2) verify the appropriate handling of degraded SSC functional performance; (3) evaluate the role of work practices and common cause problems; and (4) evaluate the handling of SSC issues reviewed under the requirements of the maintenance rule, 10 CFR Part 50, Appendix B, and the TS's. *Low pressure core spray (E21)
- Standby service water pump house ventilation (Y47)
Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted two samples.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13).1Risk Assessment and Management of Risk
a. Inspection Scope
The inspectors reviewed the seven listed assessment activities to verify: (1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and licensee procedures prior to changes in plant configuration for maintenance activities
-10-and plant operations; (2) the accuracy, adequacy, and completeness of the informationconsidered in the risk assessment; (3) that the licensee recognized, and/or entered asapplicable, the appropriate licensee-established risk category according to the risk assessment results and licensee procedures; and (4) that the licensee-identified and corrected problems related to maintenance risk assessments.*WO 77656, Low pressure core spray planned system outage*WO 77940, Installation of temporary power at Radial Well #3
- WO 51014535, Division II switchgear room cooler acid flush and cleaning
- WO 51026143, Alternate rod insertion quarterly functional test
- WO 63263, Division II EDG planned system outage*WO 51015046, Balance of plant Transformer 12B planned outage
- WO 83216, Reactor vessel pressure high annunciator troubleshooting Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted seven samples.
b. Findings
Introduction.
The inspectors identified two examples of a noncited violation of10 CFR 50.65, "Maintenance Rule," for failing to include maintenance that couldincrease the likelihood of an initiating event in the plant risk assessment.Description. The inspectors identified two instances where the licensee did not performan adequate risk assessment for plant conditions. Specifically:*On February 2, 2006, maintenance technicians conducted quarterly surveillanceProcedure 06-IC-1B21-Q-1012, "ATWS - Reactor Vessel Level / Reactor Pressure Functional Test," Revision 102, that resulted in the expected actuation of one half of the alternate rod insertion circuitry. A full actuation of the alternate rod insertion would result in a reactor scram. The inspectors noted that the work activity had not been identified as a risk activity and was not included in the licensee's assessment of plant risk. The inspectors concluded that the licensee's risk assessment was inadequate since it did not consider the increased likelihood of a reactor shutdown from the loss of redundancy in the alternate rod insertion circuitry resulting from the surveillance activity.*On March 28, 2006, maintenance technicians from the station's "fix-it-now" teamperformed emergent maintenance on a reactor vessel pressure switch which inserted in a scram signal in one division of the reactor protection system. Actuation of both divisions of the reactor protection system would result in areactor shutdown through a reactor scram. The inspectors determined that,
-11-although the control room operators were aware of the maintenance andexpected the half scram, no risk assessment had been performed for the maintenance activity.
Analysis.
The failure to include maintenance activities as part of an assessment of plantrisk is a performance deficiency that affected the Initiating Events cornerstone. Per Appendix B of Inspection Manual Chapter 0612, the finding is greater than minor since the maintenance that was performed increased the likelihood of an initiating event.
Using Inspection Manual Chapter 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process," the finding is of very low safety significance (Green) since in both cases the change in incremental core damage probability and incremental large early release probability were less than 1E-6 and 1E-7,respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnel error.Enforcement. 10 CFR 50.65(a)(4) states, in part, that before performing maintenanceactivities, the licensee shall assess and manage the risk that may result from the proposed maintenance activities. Contrary to the above, on February 2, 2006, and again on March 28, 2006, the licensee failed to adequately assess the risk associatedwith maintenance activities on alternate rod insertion and the reactor protection system,respectively. Because this finding is of very low safety significance and has been entered in the licensee's CAP as CR-GGN-2006-1041 and CR-GGN-2006-1277, this violation is being treated as a noncited violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000416/2006002-02, Failure to Perform anAdequate Risk Assessment..2Emergent Work Control
a. Inspection Scope
For the work activity listed below, the inspectors: (1) verified that the licenseeperformed actions to minimize the probability of initiating events and maintained thefunctional capability of mitigating systems and barrier integrity systems; (2) verified thatemergent work-related activities such as troubleshooting, work planning/scheduling, establishing plant conditions, aligning equipment, tagging, temporary modifications, and equipment restoration did not place the plant in an unacceptable configuration; and (3)reviewed the UFSAR to determine if the licensee identified and corrected risk assessment and emergent work control problems. *January 18, 2006, plant service water leak at radial Well 3 during excavationactivities (CR-GGN-2006-00219)Documents reviewed by the inspectors are listed in the attachment. The inspectorscompleted one sample.
-12-
b. Findings
Introduction.
The inspectors reviewed a Green self-revealing finding for a failure tofollow procedure that resulted in a significant plant service water header leak.Description. The plant service water (PSW) system at (GGNS) supplies cooling waterfor various nonessential heat loads throughout the plant. The system is supplied withwell water from four wells containing two pumps each. Though none of the components in the PSW system are safety-related, a loss of service water would result in a reactor scram.On January 18, 2006, a contractor was digging a trench to install temporary powercables in the vicinity of Radial Well #3 when he severed a 3/4-inch test connection attached to the PSW header. Lacking a means to contact the control room, the contractor entered the resident inspector office and asked the inspectors if they could turn off the service water system. The inspectors notified the control room operatorswho began monitoring service water header pressure and dispatched operators to assess the condition. Service water header pressure fell approximately 2 pounds per square inch before stabilizing.The presence of the test connection on the service water header was not indicated onthe excavation permit. The licensee determined that Drawing C1745D, "Plant Service Water System Supply and Discharge Line Plan and Profile," Revision 20, which showed the presence of the test connection, was not reviewed during preparation of the excavation permit. A pre-excavation survey of the area with radio-detection equipment also failed to identify the presence of the connection.Section 5 of Procedure EN-IS-112, "Trenching, Excavation and Ground PenetratingActivities," Revision 2, required the responsible engineer to review documents to identify potential hazards posed by underground lines in the vicinity of the excavation. This requirement was not met since the responsible engineer did not identify the presence ofthe test connection shown on Drawing C1745D. Additionally, the Responsibilitiessection of Procedure EN-IS-112 assigned the Maintenance Department responsibility forcontrolling contract personnel and the Operations Department responsibility formaintaining a continuing knowledge of the status of excavation activities. The inspectors concluded that these requirements had not been met, since the contractorinvolved had neither the means nor the knowledge of how to contact the control room after striking the test connection.Analysis. The performance deficiency associated with this finding was a failure to followthe requirements of Procedure EN-IS-112, "Trenching, Excavation, and Ground Penetrating Activities," Revision 2, resulting in the severing of a service water test connection. This finding was more than minor since it was associated with the human performance attribute of the Initiating Events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on themagnitude of the pressure reduction resulting from the line break, the inspectors determined that the mitigation functions of the plant service water system would nothave been affected. The finding was of very low safety significance (Green) since it did
-13-not contribute to the likelihood of a loss of coolant accident, did not contribute to a lossof mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. This finding had human performance crosscutting aspects associated with afailure to follow procedure.Enforcement. No violation of regulatory requirements occurred. The finding did notrepresent a noncompliance since it occurred on nonsafety-related equipment. This finding was entered into the licensee's corrective action program as CR-GGN-2006-0219 and is identified as FIN 05000416/2006002-03, Plant Service Water Leak During Excavation.
1R14 Personnel Performance During Nonroutine Plant Evolutions (71111.14)
a. Inspection Scope
The inspectors reviewed operator response to one nonroutine event during theinspection period. In addition to direct observation of operator performance, the inspectors reviewed procedural requirements, operator logs, and plant computer data to determine whether the response was in accordance with plant procedures and training.
The following event was reviewed:*On March 23, 2006, the inspectors reviewed control room personnel response toa feedwater pump trip. The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors selected six operability evaluations performed by the licensee during thereport period involving risk-significant SSCs. The inspectors: (1) reviewed plants status documents such as operator shift logs, emergent work documentation, deferred modifications, and standing orders to determine if an operability evaluation waswarranted for degraded components; (2) referred to the UFSAR and design basis documents to review the technical adequacy of licensee operability evaluations;(3) evaluated compensatory measures associated with operability evaluations;(4) determined degraded component impact on any TS; (5) used the Significance Determination Process to evaluate the risk significance of degraded or inoperable equipment; and (6) verified that the licensee had identified and implemented appropriate corrective actions associated with degraded components.*CR-GGN-2006-00007, Division I EDG load sequencing system toggle switch notin the "normal" position
-14-*CR-GGN-2006-00079, high pressure core spray pump lower motor bearing highwater content in oil sample results*CR-GGN-2006-00328, Division II EDG governor hydraulic fluid leak
- CR-GGN-2006-00467, Inadequate testing frequency for two standby gastreatment system containment isolation valves*CR-GGN-2006-00587, Standby gas treatment system failed to maintain therequired negative pressure in the auxiliary building*CR-GGN-2006-00867, Condensate and refueling water storage and transfersystem containment isolation valve did not meet closing stroke timerequirementsDocuments reviewed by the inspectors are listed in the attachment. The inspectorscompleted six samples.
b. Findings
No findings of significance were identified.
1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors selected the six listed postmaintenance test activities of risk significantsystems or components. For each item, the inspectors: (1) reviewed the applicablelicensing basis and/or design-basis documents to determine the safety functions; (2) evaluated the safety functions that may have been affected by the maintenance activity; and (3) reviewed the test procedure to ensure it adequately tested the safety function that may have been affected. The inspectors either witnessed or reviewed test data to verify that acceptance criteria were met, plant impacts were evaluated, test equipment was calibrated, procedures were followed, jumpers were properly controlled, test data results were complete and accurate, test equipment was removed, the systemwas properly realigned, and deficiencies during testing were documented. The inspectors also reviewed the UFSAR to determine if the licensee identified and corrected problems related to postmaintenance testing. *WO 63535 - Repair and retest of Division II hydrogen igniter
- WO 70065 - Standby service water inlet to the EDG jacket water cooler valveseat leakage*WO 51019608 - Upper containment airlock door inspection, hydraulic fluidchange, and maintenance retest
-15-*WO 81083 - Performance of Standby Gas Treatment A vacuum testing*WO 80889 - Division II EDG governor leakage postmaintenance test
- WO 68713 - Reactor core isolation cooling overspeed trip test The inspectors completed six samples.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the UFSAR, procedure requirements, and TSs to ensure thatthe six listed surveillance activities demonstrated that the SSCs tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the following significant surveillance test attributes were adequate: (1) preconditioning; (2) evaluation of testing impact on the plant; (3) acceptance criteria; (4) test equipment; (5) procedures; (6) jumper/lifted lead controls; (7) test data; (8) testing frequency and method demonstrated TS operability;(9) test equipment removal; (10) restoration of plant systems; (11) fulfillment of ASMECode requirements; (12) updating of performance indicator (PI) data; (13) engineering evaluations, root causes, and bases for returning tested SSCs not meeting the test acceptance criteria were correct; (14) reference setting data; and (15) annunciator and alarm setpoints. The inspectors also verified that the licensee identified and implemented any needed corrective actions associated with the surveillance testing. *January 25, 2006, Division II EDG 18-month surveillance test per Procedure 06-OP-1P75-R-0004, "Standby Diesel Generator Functional Test," Revision 111*February 6, 2006, Reactor Heat Removal 'C' valve inservice test per Procedure06-OP-1E12-Q-0007, "LPCI/RHR Subsystem C MOV Functional Test,"Revision 102*February 8, 2006, Daily calculation of identified and unidentified drywell leakageper Procedure 06-OP-1000-D-0001, "Daily Operating Logs," Revision 119*February 17, 2006, Secondary containment drawdown test per Procedure 06-OP-1T48-R-0002, "Standby Gas Treatment A Logic and Vacuum Test,"
Revision 108*February 28, 2006, Division I EDG monthly surveillance test per Procedure 06-OP-1P75-M-0001, "Standby Diesel Generator Functional Test," Revision 126
-16-*March 7, 2006, Containment ventilation and cooling system containment isolationValve M41F034 local leak rate testing per Procedure 06-ME-1M61-V-0001,
"Local Leak Rate Test Low Flow Air," Revision 108 The inspectors completed six samples.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications (71111.23)
a. Inspection Scope
The inspectors reviewed the UFSAR, plant drawings, procedure requirements, and TSsto ensure that the two below listed temporary modifications were properly implemented.
The inspectors: (1) verified that the modifications did not have an affect on system operability/availability; (2) verified that the installation was consistent with modificationdocuments; (3) ensured that the postinstallation test results were satisfactory and that the impact of the temporary modifications on permanently installed SSCs were supported by the test; (4) verified that the modifications were identified on control roomdrawings and that appropriate identification tags were placed on the affected drawings;and (5) verified that appropriate safety evaluations were completed. The inspectors verified that the licensee identified and implemented any needed corrective actions associated with temporary modifications. *February 10, 2006, Leading edge flow meter software modification perTemporary Alteration 2006-01*March 2, 2006, Secondary containment isolation Valve B21F114 was removedfrom the inputs to the Division I auxilliary building valves operable annunciatorper Temporary Alteration 2006-02The inspectors completed two samples.
b. Findings
No findings of significance were identified.
Cornerstone:
Emergency Preparedness1EP6Drill Evaluation (71114.06)
a. Inspection Scope
For the below listed drill contributing to Drill/Exercise Performance and emergencyresponse organization PIs, the inspectors: (1) observed the training evolution to assess classification, notification, and Protective Action Requirement development activities;
-17-(2) compared identified weaknesses and deficiencies against licensee identified findingsto determine whether the licensee is properly identifying failures; and (3) determined whether licensee performance is in accordance with the guidance of the Nuclear Energy Institute (NEI) 99-02, "Voluntary Submission of Performance Indicator Data,"
acceptance criteria. *January 25, 2006, the inspectors observed the licensee's emergency responseorganization in the simulator, the Emergency Operation Facility, and theTechnical Support Center respond to a simulated anticipated transient without scram event that led to fuel damage and a release to the atmosphereDocuments reviewed by the inspectors included:
- GGNS 2006 1st Quarter Emergency Preparedness Drill Evaluator's notebook*Drill Emergency Notification FormsThe inspectors completed one sample.
b. Findings
No findings of significance were identified.2.RADIATION SAFETYCornerstone: Occupational Radiation Safety [OS] 2OS1Access Control to Radiologically Significant Areas (71121.01)
a. Inspection Scope
This area was inspected to assess the licensee's performance in implementing physicaland administrative controls for airborne radioactivity areas, radiation areas, high radiation areas (HRAs), and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the TSs, and the licensee's procedures required by TS as criteria for determining compliance. During the inspection, the inspector interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspector performed independent radiation dose rate measurements and reviewed the following items:*PI events and associated documentation packages reported by the licensee inthe Occupational Radiation Safety Cornerstone*Controls (surveys, posting, and barricades) of radiation, HRA, or airborneradioactivity areas *Conformity of electronic personal dosimeter alarm setpoints with surveyindications and plant policy; workers' knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms
-18-*Physical and programmatic controls for highly activated or contaminatedmaterials (nonfuel) stored within spent fuel and other storage pools*Self-assessments, audits, licensee event reports, and special reports related tothe access control program since the last inspection*Corrective action documents related to access controls
- Licensee actions in cases of repetitive deficiencies or significant individualdeficiencies*Controls for special areas that have the potential to become very high radiationareas during certain plant operations*Posting and locking of entrances to all accessible high dose rate - HRAs and veryHRAs.The inspector completed 12 of the required 21 samples.
b. Findings
No findings of significance were identified.2OS2ALARA Planning and Controls (71121.02)
a. Inspection Scope
The inspector assessed licensee performance with respect to maintaining individual andcollective radiation exposures as low as is reasonably achievable (ALARA). The inspector used the requirements in 10 CFR Part 20 and the licensee's procedures required by TS as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed:*Current 3-year rolling average collective exposure
- Five work activities from previous work history data which resulted in the highestpersonnel collective exposures*Site-specific trends in collective exposures, plant historical data, and source-termmeasurements*ALARA work activity evaluations, exposure estimates, and exposure mitigationrequirements*Intended versus actual work activity doses and the reasons for anyinconsistencies
-19-*Records detailing the historical trends and current status of tracked plant sourceterms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry *Source-term control strategy or justifications for not pursuing such exposurereduction initiatives*Specific sources identified by the licensee for exposure reduction actions andpriorities established for these actions, and results achieved against since the last refueling cycle*Self-assessments, audits, and special reports related to the ALARA programsince the last inspectionThe inspector completed 7 of the required 15 samples and 2 of the optional samples.
b. Findings
No findings of significance were identified.4.OTHER ACTIVITIES
4OA1 Performance Indicator Verification
a. Inspection Scope
Initiating Events CornerstoneUnplanned Scrams Per 7,000 Critical HoursUnplanned Scrams With Loss Of Normal Heat RemovalUnplanned Power Changes Per 7,000 Critical HoursBarrier Integrity CornerstoneReactor Coolant System LeakageThe inspectors sampled licensee submittals for the four PIs listed above for the periodfrom January 2004 through December 2005. The definitions and guidance of NEI 99-02,
"Regulatory Assessment Indicator Guideline," Revision 3, were used to verify the licensee's basis for reporting each data element in order to verify the accuracy of PI data reported during the assessment period. The inspectors reviewed operator log entries,daily shift manager reports, plant computer data, condition reports, work orders, maintenance rule data, and PI data sheets to determine whether the licensee adequatelyreported the PIs listed above. Also, the inspectors interviewed the licensee personnel that were accountable for collecting and evaluating the PI data.
-20-Occupational Radiation Safety Cornerstone*Occupational Exposure Control EffectivenessThe health physics inspector reviewed licensee documents from April 1, 2005, throughMarch 30, 2006. The review included corrective action documentation that identifiedoccurrences in locked HRAs (as defined in the licensee's TS), very HRAs (as defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in NEI 99-02).
Additional records reviewed included ALARA records and whole body counts of selected individual exposures. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the PI data. In addition, the inspector touredplant areas to verify that HRA, locked HRA, and very HRAs were properly controlled. PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 3, were used to verify the basis in reporting for each data element.The inspector completed the required sample (one) in this cornerstone.
Public Radiation Safety Cornerstone*Radiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences The inspector reviewed licensee documents from April 1, 2005, through March 30, 2006. Licensee records reviewed included corrective action documentation that identified occurrences for liquid or gaseous effluent releases that exceeded PI thresholds and those reported to the NRC. The inspector interviewed licensee personnel that wereaccountable for collecting and evaluating the PI data. PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 3, wereused to verify the basis in reporting for each data element.The inspector completed the required sample (one) in this cornerstone.
b. Findings
No findings of significance were identified.4OA2Identification and Resolution of Problems (71152).1Routine Review of Identification and Resolution of Problems
a. Inspection Scope
The inspectors performed a daily screening of items entered into the licensee's CAP. This assessment was accomplished by reviewing work orders and condition reports and attending corrective action review and work control meetings. The inspectors: (1)verified that equipment, human performance, and program issues were being identified
-21-by the licensee at an appropriate threshold and that the issues were entered into theCAP; (2) verified that corrective actions were commensurate with the significance of the issue; and (3) identified conditions that might warrant additional follow-up through otherbaseline inspection procedures.
b. Findings and Observations
No findings of significance were identified..2Selected Issue Follow-up Inspection
a. Inspection Scope
In addition to the routine review, the inspectors selected the two listed issues for a morein-depth review. The inspectors considered the following during the review of the licensee's actions: (1) complete and accurate identification of the problem in a timely manner; (2) evaluation and disposition of operability/reportability issues; (3) considerationof extent of condition, generic implications, common cause, and previous occurrences; (4) classification and prioritization of the resolution of the problem; (5) identification of root and contributing causes of the problem; (6) identification of corrective actions; and (7) completion of corrective actions in a timely manner. January 3, 2006, Inaccurate condensate storage tank level instrumentation January 23, 2006, Klockner-Moeller valve control contact failure
b. Findings and Observations
Introduction.
The inspectors identified a Green noncited violation of 10 CFR Part 50,Appendix B, Criterion XVI, for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation.Description. While reviewing a TS amendment request, the inspectors identified anonconforming condition due to improper design of the condensate storage tank level instrumentation. In addition to this, the inspectors determined that the licenseesubstituted a manual operator action for an automatic function in which reactor core isolation cooling and high pressure core spray transfer suction to the suppression pool on low condensate storage tank level. This manual action had been in place for more than 6 years. On April 30, 1999, Condition Report CR-GGN-1999-00481 stated that, in the event offailure of the nonsafety-related portions of the condensate storage tank piping, thecondensate storage tank level transmitters would indicate an inaccurate level in the nonconservative direction; i.e., higher than actual. Section 5.4.1 of the Grand Gulf Safety Evaluation Report states, "Since the condensate storage tank is not a seismic Category 1structure, an automatic safety-grade suction switchover to the suppression pool has been provided to ensure a water supply in the event of a safe shutdown earthquake and concomitant failure of the condensate storage tank." In addition to this, the Grand Gulf
-22-Safety Analysis Report states the following in Section 6.3.2.2.1, "When the systemsenses a low water level in the condensate storage tank, the HPCS pump suction automatically transfers from this tank to the suppression pool." The condensate storage tank level inaccuracy would have prevented the automatic suction transfer function from occurring in a timely manner, which could damage both the reactor core isolation cooling and high pressure core spray pumps due to air entrainment.The licensee issued a standing order and subsequently proceduralized requirements forthe control room operators to manually transfer the suction of the high pressure core spray and reactor core isolation cooling to the suppression pool in the event of condensate storage tank failure. The inspectors concluded that the licensee's action was a manual compensatory action. The deficiency should therefore have been corrected at the earliest opportunity not to exceed the next refueling outage per the guidance at thetime of Generic Letter 91-18, "Information to Licensees Regarding NRC InspectionManual Section on Resolution of Degraded and Nonconforming Conditions," and currently reflected in the Part 9900 Technical Guidance, "Operability Determinations andFunctionality Assessments for Resolution of Degraded or Non-Conforming Conditions Adverse to Quality or Safety." Contrary to this guidance, the licensee did not correct the deficiency through replacement and recalibration of the level transmitters until December 8, 2005.Analysis. The failure to promptly correct the condensate storage tank levelinstrumentation was a performance deficiency. The finding had more than minor significance because it affected the design control attribute of the mitigating systemscornerstone and affected the cornerstone objective to ensure the availability of systemsthat respond to initiating events. The finding was of very low safety significance (Green)because it was a design deficiency that did not result in a loss of operability. This findinghad crosscutting aspects associated with problem identification and resolution in thatstation personnel did not implement corrective actions in a timely manner.
Enforcement.
10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires,in part, that conditions adverse to quality be promptly corrected. Contrary to the above, the licensee failed to promptly correct the deficient condensate storage tank levelinstrumentation. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as CR-GGN-2006-1096, this violation is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000416/2006002-04, Untimely Corrective Actions Associated with Condensate Storage Tank Level instrumentation..3Occupational Radiation Safety
a. Inspection Scope
The inspector evaluated the effectiveness of the licensee's problem identification andresolution process with respect to the following inspection areas:*Access Control to Radiologically Significant Areas (Section 2OS1)*ALARA Planning and Controls (Section 2OS2)
-23-
b. Findings and Observations
No findings of significance were identified.4OA5Other Activities.1(Closed) Unresolved Item 05000416/2005008-02: EDG Local Start Procedure
a. Inspection Scope
The inspectors completed a followup inspection of an unresolved item regarding aninadequate alternative shutdown procedure for locally starting and loading an EDG during a control room evacuation due to fire with loss of offsite power. This issue had remained unresolved to determine whether one or more credible fire scenarios existed which could cause a control room evacuation or a loss of offsite power and prevent automatic starting and loading of the Division 1 EDG.
b. Findings
Introduction.
A Green noncited violation was identified for failure to have an alternativeshutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. It was determined that a crediblefire scenario existed which could require this procedure to be used in this manner, and that the issue had very low safety significance because operator experience and familiarity with performing the required response actions were adequate to overcome theprocedure deficiency.Description. On April 12, 2005, during a walkthrough of a control room evacuation, thetriennial fire protection inspection team identified that the procedure steps in System Operating Instruction 04-1-01-P75-1, "Standby Diesel Generator System," Revision 67,called for manipulation of controls in the control room in order to manually start the Division 1 EDG. The team noted that this procedure section was not specifically written for a control room evacuation, but was referenced for use following a control room fire.
This issue was unresolved for both significance and enforcement because additional technical information was needed to assess the issue. The procedure was promptly corrected to direct operators on local starting and loading of an EDG.Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30,required operators to use System Operating Instruction 04-1-01-P75-1 to locally start the Division 1 EDG in the event that offsite power was not available. However, this procedure did not provide instructions that could successfully start and load the EDG from outside the control room. Steps to close the output breaker and load the EDG werewritten to be performed from inside the control room, which would not be possible once the control room was evacuated.
-24-Using electrical schematics, the licensee was able to demonstrate that two simplemethods were available to start and load an EDG locally. The team determined through interviews that it was likely that operators would be able to complete this action, even though it was not specifically contained in the procedure.
Analysis.
Failure to have an alternative shutdown procedure to restore power following acontrol room evacuation with loss of offsite power that was independent of the controlroom was a performance deficiency. This issue was more than minor because it affected the mitigating systems cornerstone objective for the procedure quality and protectionfrom external factors attributes. A Region IV Senior Reactor Analyst made a visit to the site during the week of January 30, 2006. Through discussions with engineers and walkdowns in the plant, the Senior Reactor Analyst determined that there is a crediblefire scenario which could simultaneously cause a control room evacuation and a loss of offsite power and prevent automatic starting and loading of the Division 1 EDG. This issue was categorized as a postfire safe shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were adequateto overcome the procedure deficiency. Therefore, this issue screened as having very low safety significance (Green) in Phase 1 of the Fire Protection Significance Determination Process (Manual Chapter 0609, Appendix F).Enforcement. Grand Gulf License Condition 2.C(41) requires that the licensee shallimplement and maintain in effect all provisions of the approved Fire Protection Program as described in Revision 5 to the UFSAR, and as approved in the Safety Evaluation dated August 23, 1991. As part of the approved Fire Protection Program, the licensee committed by letter dated August 27, 1981, to implement the requirements ofSection III.G of Appendix R to 10 CFR Part 50.Section III.G.3 of this appendix coversrequirements for alternative shutdown areas, such as the control room at Grand Gulf.
Section III.L provides requirements for the performance capability of alternative shutdowncapability necessary to comply with Section III.G.3;Section III.L.3 requires that "thealternative shutdown capability shall be independent of the specific fire area(s) and shallaccommodate post-fire conditions where offsite power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Procedures shall be in effect to implement this capability." Contrary to this, inspectors determined on April 12, 2005, that Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30, and System Operating Instruction 04-1-01-P75-1, "Standby Diesel Generator System," Revision 67,were inadequate to implement this requirement. Specifically, these procedures provided operating instructions for locally starting and loading the Division 1 EDG in the event that offsite power was not available which was not independent of equipment in the specific fire area (control room). Because this violation was of very low safety significance and has been entered in the licensee's corrective action program as CR-GGN-2005-1854, this issue is being treated as an NCV in accordance with Section VI.A of the NRC Enforcement Policy: NCV 05000416/2006002-05, Inadequate Alternative ShutdownProcedure for Locally Starting and Loading an EDG.
-25-4OA6Meetings, Including ExitOn February 27, 2006, the inspector presented the results of the review of unresolveditem (URI)05000416/2005008-02 to Mr. C. Bottemiller, Manager, Plant Licensing viatelephone. The inspector confirmed that proprietary information was neither provided nor examined during the inspection.On March 30, 2006, the health physics inspector presented the inspection results toMr. W. Brian, General Manager, Plant Operations, and other members of the staff whoacknowledged the findings. The inspector confirmed that proprietary information provided or examined during the inspection was not retained.On April 11, 2006, the resident inspectors presented the inspection results toMr. G. Williams and others who acknowledged the findings. The inspectors confirmedthat proprietary information was not provided or examined during the inspection.ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- C. Abbott, Supervisor, Quality Assurance
- C. Bottemiller, Manager, Plant Licensing
- R. Bryan, General Manager, Plant Operations
- B. Bryant, Superintendent, Chemistry
- R. Collins, Manager, Operations
- D. Coulter, Licensing Specialist, Plant Licensing
- C. Ellsaesser, Manager, Planning and Scheduling
- M. Guynn, Manager, Emergency Preparedness
- E. Harris, Manager, Corrective Action and Audits
- M. Krupa, Director, Nuclear Safety Assurance
- M. Larson, Senior Licensing Engineer
- N. Mascarella, Engineer
- C. Mason, Quality Assurance Auditor
- J. Miller, Manager, Training
- J. Owens, Senior Licensing Specialist
- J. Robertson, Manager, Quality Assurance
- M. Rohrer, Manager, System Engineering
- F. Rosser, Supervisor, Radiation Protection
- R. Sumrall, Emergency Planner
- R. Tolbert, Senior Health Physicist/Chemistry Specialist, Chemistry
- G. Williams, Vice President, Operations
- D. Wiles, Director, Engineering
- D. Wilson, Supervisor, Design Engineering
- R. Wilson, Superintendent, Radiation Protection
- P. Worthington, Supervisor, Engineering
- H. Yeldell, Manager, Maintenance
NRC personnel
- W. Walker, Senior Project Engineer, Reactor Project Branch C
- R. Bywater, Senior Reactor Analyst, Region IV
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened05000416/2006002-01URIInadequate Design Control for Freeze Protection in theDiesel Generator Building Breezeway.
AttachmentA-2
Opened and Closed
05000416/2006002-02NCVFailure to Perform an Adequate Risk Assessment(Section 1R13)05000416/2006002-03FINPSW Leak During Excavation (Section 1R13)
- 05000416/2006002-04NCVUntimely Corrective Actions Associated with CondensateStorage Tank Level instrumentation (Section 4OA2)05000416/2006002-05NCVInadequate Alternative Shutdown Procedure for LocallyStarting and Loading an EDG (Section 4OA5)
Closed
- 05000416/FIN-2005008-02URIEDG Local Start Procedure (Section 4OA5)DiscussedNone
LIST OF DOCUMENTS REVIEWED
In addition to the documents noted in the inspection report, the following documents wereselected and reviewed by the inspectors to accomplish the objectives and scope of the inspection and to support any findings:
Section 1R01:
- Adverse Weather ProtectionProcedure 04-1-03-A30-1, "Cold Weather Protection," Revision 17
- Modification Package
- ER-GG-2003-0121, Revision 0
- Drawing E-0118-014, "Heat Tracing, Diesel Generator Building Sprinklers," Revision 4
- CR-GGN-2003-00227CR-GGN-2006-00022CR-GGN-2002-2250CR-GGN-2004-0032Work Orders:
- 50298987-01, 50298970-01
Section 1R04: Equipment AlignmentProcedure 04-S-01-Z51-1, "Control Room
- HVAC System," Revision 41
- System Performance Indicator - Control Room HVAC System Piping and Instrument Diagram M-0049, "Control Room Heating, Ventilation, and AirConditioning System," Revision 39
- AttachmentA-3CR-GGN-2006-00342
Section 1R05: Fire ProtectionProcedure 10-S-03-4, "Fire Protection: Control of Combustible Material," Revision 13
Procedure
- 07-S-14-12, "Fire Extinguisher Maintenance Check," Revision 30
- Grand Guld Nuclear Station Fire Pre-Plans, Revision 15
Work Orders
- 50990896-01 and 51014895-01
- Calculation
- MC-Q1X77-96023, "Evaluate Diesel Generator Building Breezeway Airflows,"Revision 0Calculation
- MC-Q1X77-96023, Supplement 1, "Determine Maximum Allowable Outside AirTemperature With the Diesel Generator Building Outside Air Fans on High Speed and the Breezeway Banners in Place," Revison 0
Section 1R13: Maintenance Risk Assessments and Emergent Work ControlProcedure
- ENS-MP-106, "Contract Management," Revision 6
Procedure
- EN-IS-112, "Trenching, Excavation and Ground Penetrating Activities," Revision 2
Procedure
- 01-S-18-6, "Risk Assessment of Maintenance Activities," Revision 3
Procedure
- 06-IC-1B21-Q-1012, "ATWS-Reactor Vessel Level / Reactor Pressure FunctionalTest," Revision 102Work Order 77940
- Modification Package
- ER-GG-2002-0343, Revision 0
Section 1R15:
- Operability EvaluationsProcedure 06-EL-1R21-M-0001, "4.16 kV Degraded Voltage Functional Test and Calibration,"Revision 103Procedure 01-S-07-27, "GGNS Lubricating Oil Sample Program," Revision 13
- Logic Diagram E-1039, "Load Shedding & Sequencing Panel 1H22-P331," Revision 8
- AttachmentA-4
Section 2OS2: ALARA Planning and Controls, Access Controls to Radiologically SignificantAreasCR-GGN-2005-03429,
- CR-GGN-2005-03594,CR-GGN-2005-04020,
- CR-GGN-2006-01290Audits and Self-Assessments02C-GGN-2005-0022
- 2C-GGN-2005-0251Radiation Work Permits 05-1001, 05-1002, 05-1012, 06-1001, 06-1002, 06-1012Procedure 01-S-02-701, "Fuel Failure Detection and Evaluation," Revision 4
Procedure
- EDC-DC-141, "Design Inputs" Revision 0
Procedure
- EN-LI-102, "Corrective Action Process," Revision 1
Procedure
- RP-110, ALARA Program, Revision 2
Procedure
- RP-105, Radiation Work Permits, Revision 7
- LBDC 2005-074, dated November 7, 2005
Section 4OA2: Identification and Resolution of ProblemsYard Piping Drawing
- M-1400, "Condensate Storage Tank and Refueling Water Storage TankArea - Unit 1," Revision 16System Piping Isometric Drawing M-13368, "Condensate Transfer System: Condensate Supplyto RCIC & HPCS Pumps," Revision 19Level settings Diagram J-1660B, "Condensate Storage Tank A002," Revision 4
Procedure
- 04-1-01-E22-1, "High Pressure Core spray System," Revision 105
- Alarm Response Instruction 04-1-02-1H13-P870, "Panel No: 1H13-P870," Revision 116
- Off-Normal Event Procedure 05-1-02-VI-2, "Hurricanes, Tornadoes, and Severe Weather,"Revision 104Off-Normal Event Procedure 05-S-02-VI-3, "Earthquake," Revision 101
- AttachmentA-5Engineering Request
- ER-GG-1999-0217, "Condensate Storage Tank Level TransmitterReplacement," Revision 0Standing Order 99-0018
- LBDC 2005-067, dated December 5, 2005
- Piping and Instrument Diagram M-1065, "Condensate &refueling Water Storage & TransferSystem," Revision 38AECM-86/0049, dated February 15, 1986
LIST OF ACRONYMS
ALARAas low as is reasonably achievableATWSanticipated transient without scram
EDGemergency diesel generator
- GGN [[]]
SGrand Gulf Nuclear Station
HRAhigh radiation area
- NE [[]]
- NR [[]]
CNuclear Regulatory Commission
PIperformance indicator
- UFSA [[]]
RUpdated Final Safety Analysis Report