IR 05000244/2003002

IR 05000244-03-002; on 03/31 - 04/17/2003; R. E. Ginna Nuclear Power Plant; Safety System Design and Performance Capability

ML031480586
Person / Time
Site:	Ginna
Issue date:	05/28/2003
From:	Doerflein L Division of Reactor Safety I
To:	Mecredy R Rochester Gas & Electric Corp
References
IR-03-002
Download: ML031480586 (19)

Text

May 28, 2003

SUBJECT:

R. E. GINNA NUCLEAR POWER PLANT NRC INSPECTION REPORT 50-244/03-002

Dear Mr. Mecredy:

On April 17, 2003, the NRC completed a team inspection at the Ginna Nuclear Power Plant.

The enclosed report documents the results of the inspection which were discussed on April 17, 2003, with Mr. John White, Maintenance Superintendent, and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety system design and performance capability of the safety injection system and the emergency diesel generators, and compliance with the Commissions rules and regulations, and with the conditions of your license. The inspection consisted of system walkdowns; examination of selected procedures, drawings, modifications, calculations, surveillance tests and maintenance records; and, interviews with site personnel.

Based on the results of the inspection, the team identified one finding of very low safety significance (Green), which was determined to involve a violation of NRC requirements.

However, because of the very low safety significance and because it was entered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV)

consistent with Section VI.A of the NRC Enforcement Policy. If you contest the NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Ginna.

Dr. Robert In accordance with 10CFR2.790 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs 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/

Lawrence T. Doerflein, Chief Systems Branch Division of Reactor Safety Docket No.:

50-244 License No.:

DPR-18 Enclosure:

Inspection Report 50-244/03-002 w/ Attachment: Supplemental Information

cc w/encl:

P. Wilkens, President, Rochester Gas and Electric P. Eddy, Electric Division, Department of Public Service, State of New York C. Donaldson, Esquire, State of New York, Department of Law N. Reynolds, Esquire W. Flynn, President, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority D. Stenger, Ballard Spahr Andrews and Ingersoll. LLP T. Wideman, Director, Wayne County Emergency Management Office M. Meisenzahl, Administrator, Monroe County, Office of Emergency Preparedness T. Judson, Central New York Citizens Awareness Network

Dr. Robert

SUMMARY OF FINDINGS

IR 05000244/03-002; on 03/31 - 04/18/2003; R. E. Ginna Nuclear Power Plant; Safety System

Design and Performance Capability.

This announced inspection was conducted by five regional inspectors and one NRC contractor.

One finding of very low safety significance (Green) was identified, which was considered to be a non-cited violation (NCV). The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion III, for failure to support the ventilation ductwork over the safety injection (SI) and containment spray (CS) pumps, as assumed in the seismic design evaluation. In addition, the required supports were not included on the design drawings associated with the ventilation for the SI and CS pumps.

The finding is greater than minor because it affects the design control attribute of the mitigating system cornerstone objective to maintain the reliability of mitigating system equipment. The finding adversely impacts the reliability of the SI pumps and CS pumps to remain functional subsequent to a postulated seismic event, since the seismic class I ductwork and supports were not installed and configured consistent with the design analysis. The finding is of very low safety significance because it involved a qualification deficiency that did not result in a loss of function and the affected pumps remained operable.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Mitigating Systems, and Barrier Integrity

==1R21 Safety System Design and Performance Capability (IP 71111.21)

a. Inspection Scope

==

The team reviewed the design and performance capability of the emergency diesel generator (EDG) and the safety injection (SI) systems. From a risk perspective, the team focused inspection activities on components and procedures that would minimize the effects of a loss-of-offsite power (LOOP) initiating event and mitigate the associated accident sequences.

For both systems, the team verified that the existing systems were in accordance with the design basis, licensing commitments and regulatory requirements, and that the design documents, such as drawings and design calculations, were correct. The documents reviewed included engineering analyses, calculations, permanent and temporary plant modifications, piping and instrumentation drawings (P&IDs), electrical schematics, instrumentation and control drawings, logic diagrams, and instrument setpoint documentation. The team interviewed system engineers, design engineers, plant operators, and work management personnel regarding the design, operation, maintenance, and overall performance health of the EDG and SI systems and the associated support systems.

The team reviewed the operating procedures and engineering design calculations for the EDG and SI support systems in order to verify that procedure actions match design analysis assumptions. The types of procedures reviewed included: system operating procedures, abnormal and emergency operating procedures (APs and EOPs), alarm response procedures, and engineering design control procedures.

The team conducted walk-downs of accessible portions of the systems, and the associated safeguards switchgear systems, to verify the systems were consistent with design documents, calculations, and assumptions. The team used the updated final safety analysis report (UFSAR), technical specifications (TS), P&IDs, and isometric drawings as references during the walk-downs to verify the physical installation was consistent with design bases assumptions for major components, including piping, piping supports, pumps, valves, generator, and circuit breakers. During field walkdowns, the team examined the material condition of the systems, and the physical line-up of the major components. The team also walked down supporting systems including the residual heat removal system (RHR), and direct current

(dc) power supplies.

The mechanical design review focused on the capability of the EDG and SI systems, including associated supporting pumps, tanks, valves, and piping under the design basis and transient conditions. Additionally, the current performance and test criteria for the EDGs and the SI pumps and accumulators were reviewed to ensure consistency between allowable component performance and minimum allowable capabilities assumed in the accident analyses and associated design basis calculations.

The electrical design review focused on the capability of the EDGs to supply the safeguards busses and the ability of associated actuation, control, and instrumentation systems to support the design basis. The team reviewed one-line diagrams, elementary diagrams, control schematics, steady state and dynamic loading calculations, sequencer timing calculations, and protective device setpoints. The review included related operating instructions, instrument calibrations, and surveillance test procedures.

The team assessed the reliability and unavailability performance of the EDG and SI systems by reviewing selected corrective and preventive maintenance work orders (WOs) over the past two years. The team also used the Maintenance Rule System Performance Quarterly Reports and discussions with the system engineers to understand system reliability and availability. The team reviewed post-maintenance testing results for various WOs to verify the demonstrated capability of the components to perform their intended safety function.

The team also reviewed modifications to the SI and EDG support systems as well as changes to the license, TSs or plant design that could impact the functionality or reliability of the systems.

The team reviewed performance data acquired during EDG and SI TS surveillance testing (ST) activities to verify that the results demonstrated functional capability and met the acceptance criteria. Selected component performance data was reviewed to verify that test results reflected design conditions. The team observed portions of the monthly EDG STs from the field. Test acceptance criteria were reviewed and compared to design calculations, TS requirements, and inservice testing (IST)requirements. Surveillance test acceptance criteria and component online performance data results were compared with design limits to determine if the design margins were being maintained and components were properly monitored.

The team reviewed operator actions in normal procedures, APs and EOPs for operating, monitoring, and controlling the EDG and SI systems. This included a review of the adequacy of the active and passive portions of the SI system, including the injection phase and the recirculation phase. The team verified that normal, abnormal, and EOPs were consistent with systems design bases. System interfaces (instrumentation, controls, and alarms) were reviewed to assess the support to operator decision making.

The team also reviewed the ability to respond to anomalous conditions and complete recovery activities.

b. Findings

Introduction:

The team identified the licensee had failed to support the ventilation ductwork over the SI and containment spray (CS) pumps as assumed in the seismic design evaluation. In addition, the required supports were not included on the design drawings associated with the ventilation for the SI and CS pumps. This finding was determined to be of very low safety significance (Green) and non-cited violation (NCV)of 10 CFR 50 Appendix B, Criteria III, Design Control.

Description:

During a plant walkdown, the team observed that the ventilation ductwork air hoods over each of the SI pumps were positioned in close proximity to the shaft couplings and bearing oilers. However, only the air-hood for the ventilation over the B SI pump was supported by attachments to the pump pedestal. The team observed that the A and C SI pump air-hoods were not rigidly supported, and questioned whether the air-hoods could interact with the pump components during a postulated seismic event. Also, the team observed that there was significant rust on the flanged connections where the three SI pump air-hood assemblies bolted to the ductwork. In addition, the team noticed that the air-hoods over the CS pumps were removed.

The team reviewed UFSAR Section 9.4.2.4, and the associated thermal analyses, and determined that the ductwork was classified as safety significant and designed with supports to meet Seismic Class 1 requirements. However, the design no longer credited the cooling function provided by the ductwork and the upstream cooling coils had been isolated. In reviewing the seismic design, the team determined that the analysis calculated the SI and CS ductwork stiffness assuming the air-hood assemblies were connected to the pump pedestals; but the air-hoods over the A and C SI pumps were not supported. In addition, air-hood to pedestal supports for the SI and CS pumps were not included on design drawing D-118-002. Ginna personnel could not identify if the supports had been installed previously on the SI pump air-hoods.

With regard to the two CS pumps, Ginna personnel identified that the air-hoods had been removed in 1998, under PCR 98-049, because they physically interfered with the installation of new valve actuators. While the modification package concluded that the cooling function was not required, the impact of removing these air-hoods on the seismic design was not evaluated. Additionally, drawing 33013-1869 had not been updated to reflect the air flow volume change at the CS pumps.

In response to these issues, Ginna personnel walked down the equipment and initiated ARs 2003-0762 and 2003-0804 to evaluate these conditions. The evaluations concluded that the SI ductwork stiffness was likely adequate to preclude seismic interactions with pump components, and that the guards over the SI pump couplings should prevent damage to the shaft and coupling. Ginna personnel also concluded that while the CS and SI pump ductwork may deform, it would not fail during a postulated seismic event. Considering the air-hood and ductwork weight, the evaluations concluded that, notwithstanding the surface rust, the bolted joints have sufficient margin to maintain the air-hoods in place during a postulated seismic event. The team walked down the pump and duct configuration and confirmed the shaft coupling shields provided significant protection.

At the end of the inspection, Ginna personnel were preparing a plant modification (PCR 2003-010) to add support braces to the SI pump air-hoods and the CS pump ductwork to anchor them consistent with the seismic design analysis.

Analysis:

The finding adversely impacts the reliability of the A and C SI pumps and both CS pumps to remain functional subsequent to a postulated seismic event, since the seismic Class I ductwork air-hoods and supports were not installed and configured consistent with the design analysis. The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone to maintain the reliability of affected SI and CS pumps. However, the issue was determined to have very low safety significance (Green) using Phase One of the NRC Significance Determination Process (SDP) for At-Power Situations because the finding involved a qualification deficiency that did not result in a loss of function and the affected SI and CS pumps remained operable.

Enforcement:

10 CFR Appendix B, Criteria III, Design Control, requires, in part, that measures be established to correctly translate the design basis of safety related equipment, including seismic Class I equipment, into drawings and instructions; further, it requires that design control measures provide for verifying or checking the adequacy of the design. Contrary to these requirements, the air-hood supports required by the seismic design for the A and C SI pumps and the CS pumps were not shown on drawing D-118-002, likely since initial plant start-up, and were not installed in the plant.

Secondly, in 1998, modification package PCR 98-049 did not evaluate the impact on the seismic analysis of removing the air-hoods over the CS pumps, specifically, the credited connection to the floor. However, because of the very low safety significance of the issues, and because they were entered into the Ginna corrective action program (ARs 2003-0762 and 2003-0804), these issues are being treated as a non-cited violation consistent with Section VI.A of the NRC Enforcement Policy. (NCV 50-244/03-02-01)

Failure to Maintain the Ventilation Over the SI and CS Pumps in Accordance with the Design Basis

4OA2 Identification and Resolution of Problems

a. Inspection Scope

The team reviewed a sample of Action Reports (ARs) associated with the EDG and SI systems, as identified in the Attachment, to assess if Ginna personnel were identifying issues at an appropriate threshold, entering them in the corrective action program, and taking appropriate corrective actions commensurate with the significance of the issue.

The team also evaluated the basis for operability determinations resulting from the ARs.

The team also reviewed a sample of quality assurance audits and reports in the area of engineering to determine if corrective actions have been entered into the corrective action program, and if the actions were completed to resolve identified deficiencies.

Additionally, the team reviewed corrective actions for selected issues identified during two earlier NRC inspections - the previous safety system design inspection (IR 2001-05), and an electrical distribution system functional inspection (IR 1991-80).

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

On April 17, 2003, the team presented the inspection results to Mr. John Smith, Maintenance Superintendent, and other members of the licensees staff. The inspectors verified the inspection report does not contain proprietary information.

Supplemental Information Key Points of Contact Items Opened, Closed, and Discussed Abbreviations Used Documents Reviewed

-i-

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

J.

Banke

Boric Acid Corrosion Monitoring Program Coordinator

L. Berthiaume

System Engineer, Safety Injection

K. Blackall

System Engineer, Emergency Diesel Generators

M. Fitzsimmons

Analysis Engineer

M. Flaherty

Manager, Nuclear Licensing and Safety

B. Flynn

Manager, Primary Reactor Engineering

D. Gomez

Shift Supervisor

T. Harding

Nuclear Licensing and Safety Engineer

G. Hermes

Manager, Reliability (acting)

B. Hunn

Design Engineer, Electrical

G. Joss

IST Coordinator

T. Miller

System Engineer, Electrical Systems

J.

Pacher

Manager, I&C/Electrical Engineering

F. Puddu

Operating Experience Analyst

W. Rapin

System Engineer, Reactor Systems

J.

Smith

Superintendent, Maintenance

W. Tono

Analysis Engineer

C. Vitali

System Engineer

J.

Widay

Plant Manager

J.

Zapetis

Reliability Engineer

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened &

Closed

50-244/2003-02-01 NCV Failure to Maintain the Ventilation over the SI and CS Pumps in Accordance with the Design Basis LIST OF ACRONYMS AP Abnormal Procedure AR Action Report CFR Code of Federal Regulations CS Containment Spray dc Direct Current ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EOP Emergency Operating Procedure IP Inspection Procedure (NRC)

IST Inservice Test LOOP Loss of Offsite Power MDCN Modification Design Change Notice

-ii-

NCV Non Cited Violation NRC Nuclear Regulatory Commission P&ID Piping and Instrumentation Drawing PCR Plant Change Record RHR Residual Heat Removal RWST Refueling Water Storage Tank SDP Significance Determination Process SI Safety Injection ST Surveillance Test TS Technical Specification UFSAR Updated Final Safety Evaluation Report WO Work Order