ML20058G101

Final Deficiency Rept Re Misinterpretation of Design Documents & Design Error in Wiring Between Control Rod Cabinets & Core Protection Calculator.Initially Reported on 820623.Documents to Be Reviewed for Clarity

ML20058G101
Person / Time
Site:	San Onofre
Issue date:	07/19/1982
From:	Papay L SOUTHERN CALIFORNIA EDISON CO.
To:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
10CFR-050.55E, 10CFR-50.55E, NUDOCS 8208030206
Download: ML20058G101 (4)

Text

-

Ib Docket'No. 50-362

50. 55 (e) Report 7 gg 79 Southem Califomia Edison Company

,. h5 R O. BOX 800 -

2244 WALNUT GROVE AVENUE RCSEM EAD, CALIFORNIA 91770

,p ,

m.....-' July 19, 1982 ='=-=>= '*7*

Mr. R. H. Engelken, Regional Administrator U. S. Nuclear Regulatory Commission Region V 1450 Maria Lane, Suite 210 Walnut Creck, California 94596-5368

Subject:

Docket No. 50-362 San Onofre Nuclear Generating Station, Unit 3 In a letter to your office dated June 23, 1982 we identified a condition which we consider reportable in accordance with 10CFR50.55 (e) . The condition involves a misinterpretation of design documents and a design error in the wiring between control rod drive cabinets and the core protection calculator.

Enclosed in accordance with 10CFR50.55 (e) are twenty-five (25) copies of a Final Report entitled, " FINAL REPORT ON CORE PROTECTION CALCULATOR, SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3".

If you have any questions regarding this report, we would be pleased to discuss this matter with you at your convenience.

Very truly yours, Enclosures cc: R. C. DeYoung (NRC, Director I&E)

A. E. Chaffee (NRC, San Onofre Units 2 and 3) 8208030206 820719 IEFj gDRADOCK 050003gg l

^ - --. .

.f Page 1 July 19, 1982 FINAL REPORT ON CORE PROTECTION CALCULATOR San Onofre Nuclear Generating Station, Unit 3 INTRODUCTION This report is submitted pursuant to 10CFR50.55(e).

It describes a condition exhibited by the Core Protection Calculator at San Onofre Unit 2 which is applicable to Unit 3.

This report includes a description of the deficiency, an analysis of the safety implications, and the corrective action taken. By letter dated June 23, 1982, Edison con-firmed notification to the NRC of this reportable condition.

BACKGROUND During performance of post-core hot functional testing, discrepancies were discovered involving Reactor Coolant Pump (RCP) shaft speed and Control Element Assembly (CEA) position indication. These Protection Calculators (CPC' s)parameters are utilized in calculation by Core of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD).

Complete background on this condition has been given in letter from H. B. Ray to NRC, dated July 2, 1982, Subj ect:

Licensee Event Report No.82-034, San Onofrc Nuclear Generating Station, Unit 2.

DISCUSSION The following discussion is responsive to 10CFR50.55(e)(3).

The problems involving RCP shaft speed and CEA inputs are addressed separately because the causes and corrective action for each are unrelated.

, - , , - , - + - - - - .+ -

< 1 1,

Page 2

1. RCP SHAFT SPEED INPUTS _

A. Description of Deficiency During the post core hot functional testing, a Unit 2 l reactor operator reported that the CPC point identification (PID) I for the RCP Shaft Speed displayed at the CPC operator's console did not I correspond to actual pump inputs. This problem was caused by a misinterpretation by the reactor operator regarding pump numbering schemes. Investigation has confirmed that the originally wired RCP speed input is the correct configuration.

B. Analysis of Safety Implications As explained in the previously referenced Licensee Event Report, the original RCP wiring was correct. Furthermore.

had the wiring changes which were suggested by the misinterpre-tation been made, there would have been no adverse effect on safety.

C. Corrective Actions Design documents related to the misinterpretation will be reviewed to determine whether clarifications or changes are required.

II. CEA POSITION INDICATION INPUTS A. Description of Deficiency.

The CEA position indication problem was identified during Control Element Drive Mechanism (CEDM) Rod Drop Testing.

It was noticed that the CPC operator's console CEA assignments did not correspond to actual CEA assignments. It was determined that the as-built wiring of the CEA inputs was consistent with applicable C-E and vendor drawings, but the CPC software did not correspond to the hardware documentation. Investigation of the problem indicated that, while adequate quality assurance pro-cedures existed to implement and verify the CEA input order, an error was nade in the implementation of the process.

r-

. .s

page 3 B, Analysis of Safety Implication If the CEA assignment problem had remained undetected and the CPC had been in service, response would have been con-servative. As explained in the previously referenced Licensee Event Report, the response would have resulted in a reactor trip or alerting the operators to the problem. A review of.the mechanism-i of the problem indicates that it is not subject to propagation into the areas of the CPCS software design.

C. Corrective Action

1. A design change was issued on June 23, 1982.to revise the CEA inputs to make them compatible with the existing CPC software. This design change includes conducting verification checks of the entire circuit at the completion of the rewiring to-verify the correct pin assignments.

2. The supplier of the system is developing a Software / Hardware Interface Specification which will define the plant locationof identified sensors and trace individual signal paths.

This document will be quality assured in accordance with the supplier's -

Quality Assurance procedures and will be reviewed and approved by each cogizant design group. It will then be included as part of the software design documentation. This action will be completed by August 31, 1982.

a i

820719 LLS 1

- - . , , --, . -. . _ . , .. - .- - - - . - . - , , - . , - - , . , - - . . - . . - . . . - , . . - ~ . ,