IR 05000424/1989011

Insp Rept 50-424/89-11 & 50-425/89-13 on 890307-10 & 29-31. No Violations or Deviations Noted.Major Areas Inspected: Proposed Startup Tests for Unit 2 & Followup of Previous Violation on Unit 1

ML20244E197
Person / Time
Site:	Vogtle
Issue date:	04/10/1989
From:	Belisle G, Burnett P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20244E195	List: IR 05000424/1989011 ML20244E192
References
50-424-89-11, 50-425-89-13, NUDOCS 8904240344
Download: ML20244E197 (6)

Text

_

.- . _ .

g < ,

, .

, , pd5d "N ,

l . p' . .. k

-

.

UNITED STATES

!

g-g j! l NUCLEAR REGULATORY COMMISSION REGION ll l A,t g 4,,,, [ 101 MARIETTA ST., ATLANTA, GEORGIA 30323

'

s Report Nos.: .50-424/89-11 and 50-425/89-13 L Licensee: Georgia Power Company P. O. Box 1295

. Birmingham, AL.35201 Docket.Nos.: 50-424 and 50-425 License Nos.: NPF-68 and NPF-7 ' Facility Name: Vogtle 1 and 2

/ Inspection Conducted: l March 7 - 10 and 29 - 31,1989 Inspector:

F' T. BurnEft jo 7

'

///N/M8M QKte ' Signed Approved by:. C[ 0 f G. A. Belisle, Chief Date Signed Test Programs Section Engineering Branch Division of Reactor Safety-SUMMARY Scope This routine, unannounced inspection addressed the areas of proposed startup tests for. Unit 2 and followup of a previous violation on Unit Results-o The proposed startup test to demonstrate remote shutdown capability .used more personnel than the licensee had represented as required for L the , task and. exceeded the minimum shift crew size as specified in I technical specifications and described in the Final Safety Analysis Report. ' Following extensive discussions with the licensee and Region lII management, a determination was made that the test must be con-

' ducted with a minimum crew of four,. of whom only one could be a senior reactor operato The licensee issued revision 1 to the procedure to conduct the test with a crew of four. (Paragraph 2)

Documentation of computer software changes has been significantly improve However, a recent software addition to the ERF computer i'

did not: reflect the most recent version of the surveillance procedure it was . supposed to emulat The solution to this user-programmer interface problem is still under review. (Paragraph 5).

No violations or deviations were identifie PDR ADOCK 05000424 g PDC a______-___________________ - - - _ _ _ _ _

- _ - .- _ _ _ _ _ _ _ - _ _

,

w ,.

y' ,

.

[- ,

. .

'

p .

REPORT DETAILS i

H 1.- Persons-Contacted

'

' Licensee Employees

• M.- J.= Ajluni, Operations Superintendent

1. • G. B. Bockhold, Jr., General Manager, Vogtle . Nuclear Operations

• R. M. Bellamy, Plant Manager

.

F. J. Ealick, Plant Computer Engineer

1. G. R. Frederick, . Quality Assurance Site Manager TH #*W. C. Gabbard, Senior Regulatory Specialist H. M. Handfinger, Maintenance Manager T. S. Hargis, On-shift Operations Supervisor G. L. Hooper, Plant Computer Engineer-

1. • W. F. Kitchens, Assistant General Manager for Operations and Maintenance

1. A. L. Mosbaugh,' Plant' Support Manager -

R. A. _Moye,: P1 ant Computer Engineering Supervisor

1. C. F. Meyer, Operations Superintenden ;W. T. Nicklin, Regulatory Compliance Supervisor

.

1. • R. M. Odom, P1 ant Engineering Supervisor

.

~ #*J. E. Swartzwelder, Operations Manager Other licensee- employees contacted included engineers and office - '

personne Other Organizations

.C. B. Holland, Westinghouse

'

W. C. Phoenix, Consul Tec-0. D. Hayes, Consul Tec NRC Resident Inspectors

1. • R. F. Aiello, Resident Inspector

• J. F. Rogge, Senior Resident Inspector, Operations

• C. W. Burger, Resident Inspector

• Attended exit interview March 10, 198 '# Attended exit interview March 31, 198 Acronyms and initialisms used throughout this report are listed in the last paragrap _ _ _ . . _ _ . _ _ _ - _ _ _ _ - _ _ - _ _ - _ . _ _ _ ___ _ . ..

__ _ _ _ _ . .__ __ _________ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ - _ _

!. .

'

.

I; .

.

.

l Power Ascension Test Procedure Review: Shutdown from Outside the Control Room (72583) References The'following references were reviewed by the inspector to determine the acceptability of the test as proposed:

.;

(1) FSAR 14.2.8.2.45, Remote Shutdown Test, (2) FSAR 9.5.1.5.3, Fire Brigade, (3) Regulatory Guide 1.68.2 (Revision 1, July 1978), Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Plants,.

(4) Preoperational Test Procedure 2-300-10, Shutdown Panel Preoper-ational Test, completed October 22, 1988, (5) Technical Specification Table 6.2-1, Minimum Shift Crew Composition,  ;

(6) Startup Test Procedure 2-600-10 (Revision 0), Remote Shutdown l Test, (7) Abnormal Operating Procedure 18038-02 (Revision 2), Operation from Remote Shutdown Panels, and (8) NRC Inspection Report No. 50-424/88-1 Discussion Acceptable tests to demonstrate the remote shutdown capability required by GDC 19, are described in reference 2.a(3). Two tests are described: the hot standoy demonstration procedure and the cold shutdown demonstration procedure. From review of reference 2.a(4),

the inspector concluded that cold shutdown capability had been adequately demonstrated and that reference 2.a(6) could be limited to the hot standby demonstratio ' Key features required in the hot standby demonstration include:

(1) initiation of the test from outside the control room (trip the reactor remotely),

(2) performance of the test with the minimum shift crew as discussed in references 2.a(2) and 2.a(5),

(3) achievement of the hot standby after the remotely initiated trip from a power level of 10 - 25% RTP, and

(4) maintenance of a stable hot standby condition for at least 30 minutes.

For two reactors in mode 4 or higher, reference 2.a(5) defines a

!

minimum shift crew of eight, or four per reactor. The specification l

does allow one of the minimum of eight to be absent for up to two hours. Hence, one unit might be served by a crew of only thre Further, in reference 2.a(2), the licensee states that fire team l manning requirements will be met without impacting the minimum

'

_ _ _ _ _ _ _ _ _ _ _ - _

._

-

L

-

! ,

,

1'

on-shift operating . requirements of a three-memoer remote shutdown team per unit. In reference 2.a(8), the results of a walk through of the Unit 1 procedure, similar to reference 2.a(7), using three people and a simulator demonstration of remote shutdown using three people were reported as successful. Nevertheless, the proposed startup test, reference 2.a(6), assigned five people to the task. In discus-sions with licensee personnel, they continued to aver that a unit could be shutdown remotely with only three people without' risk to the general public, but they resisted performing the required test, reference 2.a(1)', with only three people on the grounds that they could not protect plant equipment. Additionally, they cannot staff all three remote pancls with only three people. Reference 2.a(3)

does allow non-test personnel to perform routine activities to protect equipment provided those actions do not affect the test. Attachment 1 to reference 2.a(6) specifies permissible actions by control room observers. All actions by the control room observers must be logged on data sheet 7.2 of that reference and will be reviewed for effect on the test before the test is accepted by the license The inspector discussed the issue with Regional management, who concluded that a minimum shift crew for this test was four persons, including not more than one senior reactor operator. The rationale was that the test was to be conducted as only a single failure, loss of control room, and that the consideration of an additional failere, such as an incapacitated operator or fire, was not a test require-ment. This position was relayed to the licensee by the inspecto The licensee agreed to conduct the test with the defined minimum crew and issued revision 1 to reference 2.a(6) to implement the agreemen The four-person crew will permit staffing all three remote shutdown panel No violations or deviations were identifie . Power Ascension Test Procedure Review: Loss of Offsite Power (72582) , References The inspector reviewed the following documents:

(1) FSAR 14.2.8.2.46, Loss of Offsite Power at Greater than 10-Per-cent Power Test, (2) Regulatory Guide 1.68, Intial Test Programs for Water-Cooled Nuclear Power Plants, Appendix A, paragraph 5.jj, and (3) Startup Test Procedure 2-600-09 (Revision 0), Loss of Offsite Power at Greater Than 10% Powe The proposed test is scheduled by procedure 2-600-13, Power Ascension Test Sequence, to be performed at about 17% RTP but before exceeding 30% .

RT The test, reference 3.a(3), provides for simulation of loss of l offsite power by opening seven breakers on the low side of the reserve auxiliary transformer instead of the two high side breaker Review of power distribution drawings with the licensee confirmed

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

____ __ __ ____ __

f

.

.

,

-

.

that the seven low-side breakers were functionally equivalent to the two high-side breakers with respect to interrupting power to the unit. However, the two class 1E buses have under voltage relays that would be exercised by opening the high-side breakers, but would not be tested by opening the individual low-side bus' breakers. Review of preoperational tests with the licensee personnel who performed them convinced the inspector that the under-voltage relays had been acceptably tested and found fully functional. Hence, reference 3.a(3)

was accepted as writte No violations or deviations were identifie . Review of Completed Initial Fuel Loading Procedures (72500)

The completed test procedures listed below were reviewed following comple-tion of the licensee's review process and acceptance of the test results by plant managenient, , Initial Fuel Load Test Sequence, completed on February 17, 1989, and , Inverse Count Rate Ratio Monitoring for Fuel Loading, completed on February 17, 198 No test deficiencies or violations were identified in the revie . Followup on Items of Noncompliance (92702)

(Closed) Violation 50-424/88-06-01: Failure to have an adequate program to control computer software used in surveillance. The licensee's response, dated March 21, 1988, was found acceptable by Region II. The inspector's review of the corrective action, new and acceptable administrative proce-dures, was reported in Inspection Report No. 50-424/88-57. During this inspection, examples of software changes and the associated CSCRs on both the ERF and PROTEUS computers were inspected and discussed with the computer engineers responsible for the changes. Documentation and instal-lation testing of the changes by the computer engineers were complete and acceptable. However, one software change to provide RCS leakage monitor-ing on the ERF computer was based upon a superseded version of the sur-veillance procedure (14905-1 Revision 19)). Furthermore, the calculation of change in inventory was based upon changes in cold-leg temperature rather than system average temperature. The operations department had not yet made use of the software or compared its performance with the surveil-lance procedure. The solution to this user-programmer interface problem is under review by the licensee, and will be addressed in a future inspec-tion, j 4 L

- _ _ _ _ _ _ . ._-_ _ _ .

,.

g

- .

i

> Exit Interviews (30703)

The . inspection scope and findings were summarized on March 10 and 28,1989, with those persons indicated in paragraph 1 above. The inspector de- i scribed the areas inspected and discussed in detail the inspection find-

'

ing Dissenting comments were not received from the license Proprietary information is not contained in this report. Licensee manage-ment confirmed their commitment to perform the remote shutdown test with a crew of four, including only one senior reactor operato . Acronyms and Initialisms Used in This Report CSCR - computer software change request ERF - emergency response facility FSAR - Final Safety Analysis Report GDC - general design criterion from 10CFR50 Appendix A PROTEUS trade name for the plant computer RCS - reactor coolant system RTP - rated thermal power i

_ _ . - . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _