ML023300568

Notification of Sequoyah Nuclear Plant - Safety System Design and Performance Capability Inspection - NRC Inspection Report 50-327/2003-02 and 50-328/2003-02

ML023300568
Person / Time
Site:	Sequoyah
Issue date:	11/26/2002
From:	Ogle C Division of Reactor Safety II
To:	Scalice J Tennessee Valley Authority
References
IR-03-002
Download: ML023300568 (6)

See also: IR 05000327/2003002

Text

November 26, 2002

Tennessee Valley Authority

ATTN: Mr. J. A. Scalice

Chief Nuclear Officer and

Executive Vice President

6A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

SUBJECT: NOTIFICATION OF SEQUOYAH NUCLEAR PLANT - SAFETY SYSTEM

DESIGN AND PERFORMANCE CAPABILITY INSPECTION - NRC

INSPECTION REPORT 50-327/2003-02 and 50-328/2003-02

Dear Mr. Scalice:

The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)

Region II staff will conduct a safety system design and performance capability inspection at

your Sequoyah Nuclear Plant during the weeks of February 10, 2003, and February 24, 2003.

A team of five inspectors will perform this inspection. The inspection team will be led by

Mr. Jim Moorman, a Senior Reactor Inspector from the NRC Region II Office. This inspection

will be conducted in accordance with baseline inspection program Attachment 71111.21, Safety

System Design and Performance Capability.

The inspection will evaluate the capability of installed plant equipment to detect and respond to

a steam generator tube rupture event. Procedures which direct the mitigating actions for this

event will also be evaluated.

During a telephone conversation on November 25, 2002, Mr. Moorman of my staff, and

Mr. James Proffitt of your staff, confirmed arrangements for an information gathering site visit

and the two-week onsite inspection. The schedule is as follows:

• Information gathering visit: Week of January 5, 2003

• Onsite inspection: February 10-14, 2003 and February 24-28, 2003.

The purpose of the information gathering visit is to obtain information and documentation

outlined in the enclosure needed to support the inspection. Mr. Rudolph Bernhard, a

Region II Senior Reactor Analyst, may accompany Mr. Moorman during the information

gathering visit to review probabilistic risk assessment data and identify risk significant

components which will be examined during the inspection. Please contact Mr. Moorman prior

to preparing copies of the materials listed in the Enclosure. The inspectors will try to minimize

your administrative burden by specifically identifying only those documents required for

inspection preparation.

During the information gathering visit, the team leader will also discuss the following inspection

support administrative details: office space; specific documents requested to be made

available to the team in their office space; arrangements for site access; and the availability of

TVA 2

knowledgeable plant engineering and licensing personnel to serve as points of contact during

the inspection.

Thank you for your cooperation in this matter. If you have any questions regarding the

information requested or the inspection, please contact Mr. Moorman at (404) 562-4647 or me

at (404) 562-4605.

In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document

Room or from the Publically Available Records (PARS) component of 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/

Charles R. Ogle, Chief

Engineering Branch 1

Division of Reactor Safety

Docket Nos.: 50-327,328

License Nos.: DRP-77, DRP-79

Enclosure: Information Request for the Safety System Design and

Performance Capability Inspection

cc w/encl:

Karl W. Singer

Senior Vice President

Nuclear Operations

Tennessee Valley Authority

Electronic Mail Distribution

James E. Maddox, Acting Vice President

Engineering and Technical Services

Tennessee Valley Authority

Electronic Mail Distribution

Richard T. Purcell

Site Vice President

Sequoyah Nuclear Plant

Electronic Mail Distribution

(cc w/encl contd - See page 3)

TVA 3

(cc w/encl contd)

General Counsel

Tennessee Valley Authority

Electronic Mail Distribution

Robert J. Adney, General Manager

Nuclear Assurance

Tennessee Valley Authority

Electronic Mail Distribution

Mark J. Burzynski, Manager

Nuclear Licensing

Tennessee Valley Authority

Electronic Mail Distribution

Pedro Salas, Manager

Licensing and Industry Affairs

Sequoyah Nuclear Plant

Tennessee Valley Authority

Electronic Mail Distribution

D. L. Koehl, Plant Manager

Sequoyah Nuclear Plant

Tennessee Valley Authority

Electronic Mail Distribution

Lawrence E. Nanney, Director

TN Dept. of Environment & Conservation

Division of Radiological Health

Electronic Mail Distribution

County Executive

Hamilton County Courthouse

Chattanooga, TN 37402-2801

Ann Harris

341 Swing Loop

Rockwood, TN 37854

John D. White, Jr., Director

Tennessee Emergency Management Agency

Electronic Mail Distribution

Distribution w/encl: See page 4

TVA 4

Distribution w/encl:

R. W. Hernan, NRR

RIDSNRRDIPMLIPB

PUBLIC

OFFICE RII:DRS RII:DRP

SIGNATURE MOORMAN CAHILL

NAME JMOORMAN SCAHILL

DATE 11/26/2002 11/26/2002 12/ /2002 12/ /2002 12/ /2002 12/ /2002 12/ /2002

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

PUBLIC DOCUMENT YES NO

OFFICIAL RECORD COPY DOCUMENT NAME: C:\ORPCheckout\FileNET\ML023300568.wpd

INFORMATION REQUEST FOR THE SAFETY SYSTEM DESIGN AND

PERFORMANCE CAPABILITY INSPECTION

STEAM GENERATOR TUBE RUPTURE EVENT

Note: Electronic media is preferred if readily available. (The preferred file format is

searchable .pdf files on CDROM.)

1. Design basis documents for the engineered safety features used to mitigate the steam

generator tube rupture (SGTR) event. Design basis documents for pressurizer relief

valves, secondary system relief valves, atmospheric dump valves and turbine bypass

valves. Include performance history of these valves for the past 10 years.

2. All procedures used to implement the mitigation strategy for the SGTR event. Include

emergency, abnormal, and normal operating procedures as appropriate.

3. Procedures used for the operational testing of check valves in portions of the

emergency core cooling systems used during mitigation of the SGTR event.

4. Surveillance procedures used to ensure the operability of equipment required by your

Technical Specifications that is used during the mitigation of the SGTR event

5. Summary results of the steam generator (SG) in service inspection program. (Unit 2

only)

6. List of temporary modifications and operator work-arounds involving any components

required for detection or mitigation of a SGTR event for the past 3 years

7. System description and operator training modules for a SGTR event.

8. List of operating experience program evaluations of industry, vendor, or NRC generic

issues related to a SGTR event.

9. Procedures used to sample the reactor coolant system during a SGTR event.

10. Calibration and functional testing procedures for the main steam line radiation

monitoring instrumentation used to detect and mitigate a SGTR event.

11. Calculations used to support the set points in Emergency Operating Procedures for a

SGTR event.

12. Performance history of valves or support equipment used to isolate SGs in the event of

a tube rupture.

13. Calibration and functional test procedures of instruments used to monitor reactor coolant

system (RCS) pressure, pressurizer level and pressure, SG level and pressure, hot and

cold leg temperature, subcooling monitor, feedwater flow, steam flow, core exit

temperature, high pressure injection (HPI) flow, low pressure injection flow, refueling

water storage tank level, pressurized heater status, safety relief valve position indicator,

Enclosure

2

auxiliary feedwater flow (AFW) flow, condensate storage tank (CST) level, makeup flow,

and letdown flow.

14. P&IDs for RCS, HPI, SI, AFW, chemical and volume control system, main steam, and

letdown. (Paper copies are preferred for these)

15. Electrical schematic showing start circuit for the AFW pumps. (Paper copies are

preferred for these)

16. Test procedures for the primary and secondary system safety relief valves including any

position indications and code safety valves.

17. Loop uncertainly calculations for SG level, pressurizer pressure and level, and RCS

pressure.

18. Test procedures for any defeat switches associated with AFW starting logic.

19. Instrument loop diagrams for items identified in Number 13 above. (Paper copies are

preferred for these)

20. Probability Risk Assessment (PRA) Event tree for a SGTR event. A list of PRA

identified system dependencies and success criteria for systems used to mitigate a

steam generator tube rupture.

21. System health reports and all performance monitoring information for systems used to

detect and mitigate a SGTR event.

22. A list of Problem Evaluation Reports and non-routine work requests initiated since 1998

affecting the systems used to detect and mitigate a SGTR event.

23. Maintenance Rule performance criteria for systems used to detect and mitigate a SGTR

event. A list of maintenance rule failures of equipment used to detect or mitigate a

SGTR event.

24. Key electrical single line diagrams of the alternating current and direct current power

systems that provide power for the pumps, valves, and instrumentation and control

circuits associated with the systems that accomplish the SGTR mitigation strategy.

(Paper copies are preferred for these)

25. Provide a list of equipment used to mitigate a SGTR that changes state or is manually

manipulated during implementation of the SGTR mitigation strategy. Provide equipment

failure rates over the past 10 years for these components.

Enclosure