IR 05000247/1982010
| ML20055B572 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/24/1982 |
| From: | Caphton D, Eapen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20055B567 | List: |
| References | |
| 50-247-82-10, NUDOCS 8207220509 | |
| Download: ML20055B572 (6) | |
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^ U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
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Report No.
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50-247
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-DPR-26 Priority
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License No.
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Licensee:
Consolidated Edison
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4-Irving Place
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New York, New York 10003 Facility Name:
Indian Point 2 Ifispection At:
Indian Point 2 and Corporate Offices
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Inspection Conducted:
May 24 through 28, 1982
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d. kojmw (>!/7/N Inspectors:
P. K! Eapen, Fh.D.! Reactor Inspector
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U. L.~ Cap'litdn, Chief, Management Program
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Inspection Summary:
Inspection May 24 through 28, 1982 (Report No. 50-247/
l 82-10) Routine unannounced inspection of facility modification. The inspection involved 28 inspection hours on site and 14 inspection hours at the corporate
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offices by one_ region based inspector.
Results:
Two violations were identified in the inspected area, failure to follow procedures and inadequate design interface control.
(see paragraph 2.e)~
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i 8207220509 820702 PDR ADOCK 05000247
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DETAILS 1.
Persons Contacted M. Anderson Senior Operator Training Instructor J.' Basil General Manager of_ Nuclear Power Generation C. Carbonara Engineer, Nuclear Licensing D. Domey Director, Quality Assurance P. Duggan Principal Engineer W. Ferreira Plant QA Engineer D. Gaynor Sr. Engineer, Nuclear Systems Engineering
- G. Groscup Assistant Vice President
- R. Helfant Manager, Mechanical Program Section V. Jayaraman Sr. Engineer, C&I Subsection J. Mills Manager (Acting) Program Development S. Nadipuram Consultant, Mechanical Engineering D. Rush Principal Field Engineer M. Scott Engineer, Nuclear Licensing M. Smith Chief Technical Engineer P. Szabados Head, Electrical Generation and Controls Subsection G. Wasilenko Principal Consultant, Quality Assurance
- P. Zarakas Vice President
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The inspector also held discussions with other members of the Station and Corporate Technical and Administrative Staff.
- Denotes those present at the exit interviews conducted at the corporate office on May 27, 198.
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2.
Design Change / Modification Control a.
References 10 CFR Part 50, Appendix B
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ANSI N 45.2-1977 ANSI N 45.2.11-1974
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Regulatory Guide 1.64 (Revision 1)
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Procedure CI-240-1 (Including ACN 67 issued on April 2, 1982)
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Procedure OP-290-1 (January 4, 1982)
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Review
The design change packages listed in c., below, were reviewed on a sampling basis to verify that the following requirements have been met, as applicable:
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Design Input Requirements, such as design bases,. regulatory requirements, codes, and standards were identified, documented and their selection reviewed and approved.
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Design activities shall be prescribed and accomplished in
accordance with procedures that would assure the applicable design inputs are correctly translated into specifications, drawings, procedures, or instructions.
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Interface controls were established to identify, control and maintain responsibilities, lines of communications, and documentation requirements for interval and external interfaces.
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. Design verification was established to determine the adequacy of the design to meet the requirements specified in design
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Document control procedures were established to control the issuance of design documents and their changes.
Design change control-procedures were established to control
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design changes.
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Design documentation and records were maintained.
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Audits were conducted to verify compliance with all aspects of QA programs for design and design changes.
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New or modified systems were installed in accordance with the approved design.
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New or revised procedures relating to the modified system were completed and approved for technical specifications.
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As built drawings were revised to reflect modifications.
The operators were trained to use the modification.
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c.
Document / Record Packages
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ESG-80-2-05 PORV Direct Alarm Initiation
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ESG-80-2-13 Reactor Vessel Level Indication
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ESG-80-2-14 Containment High Pressure ESG-80-2-15 High Range Containment Radiation Monitor
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ESG-80-2-16 High Range Noble Gas Effluent Monitor
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ESG-80-2-17 Post Accident H -0 Containment Air Sampling 2 2
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ESG-80-2-23 Post accident RCS Sampling System
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ESG-80-2-28 Containment Water Level Indication
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MMC-80-2-19 MS Rad. Monitors
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MMC-80-2-13 H Recombiner S0V's
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MMC-80-2-12 Iso. Valve Seal Water System l
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MMC-80-2-11 N Purge for Gas Sampling Lines
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MMS-80-2-03 Containment Isolation Valve to MOV
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MMC-80-2-05 Containment Air Sampling MMC-80-2-08 High Range-Vent Noble Gas Monitor
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FFI-80-2-07 Reactor Coolant Seal Injection
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MMC-77-2-04 ASME Sect. XI Pump Flow
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MMC-80-2-04 High Range Containment Pressure Indication
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MMC-80-2-06 Post Accident Sampling System
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MMS-79-2-04 Reactor Coolant System Vents d.
Detailed Interviews and Examination of Certain Documents The inspector interviewed cognizant Engineering personnel both at the plant site and at the Corporate headquarters to determine the effectiveness of the licensee's design control program as they related to the Design change packages listed in item c above. For each design change, the following documents were reviewed to verify the adequacy of the licensee's engineering overview and administrative controls.
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Design Requirements Design Drawings
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Design Verification Safety Review
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e.
Findings The inspector noted:
(1) The designs by the field Engineering were generally reviewed by the originator's Supervisor.
For FFI-80-2-06 the design was reviewed and approved by the originator himself. The inspector pointed out to the licensee's representatives that the above actions were contrary to the requirements of the references listed in Section 2.a of this report and were examples for the violation discussed below:
(2) The documentation for the modification listed in Section 2.c of this report did not contain results of the independent review as required by Section 5.2 of Procedure OP-290-1.
The inspector stated that the above was an example of the violation discussed below. The licensee's representatives acknowledged the inspector's statements.
(3) The records for MMC-80-2-04 contained documentation for an unrelated design change.
These records should have been reviewed and verified in accordance with Construction Field Directive No. 8(Rev.0).
The inspector stated that this was another example for the violation discussed below.
The licensee's representatives acknowledged the inspector's statemen..
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Items (1) through (3) above are examples of failure to follow licensee's own procedures and as such they constitute a violation.
(50-247/82-10-01)
(4) Preoperational tests for MMC-80-2-04 did not contain response time measurements. This verification is required by NUREG 0737 (Item II.F.1, Attachment 4). The inspector noted that the required inforn:ation was not transmitted to the Preop Test group by the Engineering Staff.
The inspector informed the licensee's representatives that the above was an example of inadequate design interface control and constituted a violation (50-247/82-10-02). The licensee acknowledged by stating that the required response time would be estimated from component response time and justified; and the actual response time would be measured at the next outage of suitable duration.
The inspector informed the licensee's representatives that these findings should be considered as symptoms of problems in the licensee's design change control program; therefore merely correcting inspector identified items would not constitute an acceptable corrective action.
3.
Exit Interview The inspector met with the licensee's corporate managment denoted in paragraph 1 of this report on May 27, 1982 to discuss the findings in corporate program. At the conclusion of the inspection, on May 28, 1982, the inspector met with the general marager of Nuclear Power Generation to summarize the plant specific iindings.
The licensee's on site and corporate management acknowledged the inspector's summary statements.