ML17199Q923
| ML17199Q923 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/27/1987 |
| From: | Harrison J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17199Q903 | List: |
| References | |
| 50-237-87-06, 50-237-87-6, 50-249-87-11, IEB-79-14, NUDOCS 8707310039 | |
| Download: ML17199Q923 (2) | |
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NOTICE OF VIOLATION Commonwealth Edison Company Docket Nos. 50-237; 50-249 As a result of the inspection conducted from.January 14 through July 17, 1987, and in accordance with 10 CFR Part 2, Appendix C - General Statement of Po._licy
. and Procedure for NRC Enforcement Actions (1985), the following violations were identified:
- 1.
10 CFR 50, Appendix B, Criterion V, as implemented by CECo Topical Report CE-1-A,.
11Qual ity Assurance Program for Nuclear Generating Stations, 11 and CECo Corporate Quality Assurance Manual, Nuclear Generating Stations, "Quality Requirements," requires that activities affecting quality shall be prescribed by documented procedures and drawings and shall be accomplished in accordance with these procedures and drawings.
_Procedures shall include appropriate quantitative acceptance criteria for determining that activities have been satisfactorily accomplished.
Contrary to the above, certain activities were not accomplished in accordance with documented drawings and inadequate drawings and procedures were prescribed for certain activities in that:
- a.
Structural steel connections inside the drywell were not accomplished in accordance with the General Electric drawings.
Certain beam connections had missing welds, missing bolts or welds that were apparently cut so they were no longer effective. This caused nine connections to exceed the allowable design stress limits.
(237/87006-0lA; 249/87011-0lA)
- b.
Unit*2 Drywell Structural Beam R-19 was removed and reinstalled as part of a pipe repair procedure.
The reinstallation process
. mistakenly left a 1 inch by 1 inch notch in the flange of this beam.
The reinstallation was not* accomplished in accordance with the design drawings.
This caused the beam to exceed the allowable design stress limit.
(237/87006-018)
- c.
Support M-3208-08 was installed 13 inches.outside the tolerance specified for this support.
In addition, Support 1403-M-201 was not removed from the piping system as specified on Drawing M-3208-08.
(237/87006-0lC; 249/87011-018)
- d.
Procedure DTP-2 "Inservice Inspection Plan 11 Revision 2, March 1985, and DAP-11-8 "Non-Destructive Testing (In-Service Inspection),
11 Revision 3, August 1985, did not prescribe adequate instructions or acceptance criteria concerning spring can settings.
(237/87006-010; 249/87011-0lC)
- e.
Drawing ISI-200, Sheet 1 of 3, Revision A, February 15, 1980, did not include new supports, deleted supports or new support numbers that resulted from IE Bulletin 79-14 and Mark I modifications.
8707310039 870727 PDR ADOCK 05000237 G
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Without accurate drawings, the in-service inspections for component supports could not be adequately accomplished.
(237/87006-0lE; 249/87011-010).
- f.
The visual examination deficiency for Support 1403-M-201 was not promptly communicated to the engineers performing an operability analysis of the same piping system.
There were no procedures that prescribed the required corrective actions.
(237/87006-0lF; 249/87011-0lE)
This is a Severity Level IV violation (Supplement 1).
- 2.
10 CFR 50, Appendix B, Criterion III, as implemented by CECo Topical Report CE-1-A, 11Quality Assurance Program for Nuclear Generating Stations, 11 and CECo Corporate Quality Assurance Manual, Nuclear Generating Stations, "Quality Requirements, 11 requires that measures shall be establish to assure that the design bases are correctly translated into drawings.
Contrary to the above, the design bases were not correctly translated into drawings in that:
- a.
- The fabrication drawing for embedment plates incorrectly specified an 18 inch spacing for anchor straps instead of 9 inches.
The embedment plates were subsequently fabricated different from the d~sign basis causing three plates to exceed the design stress limit.
(237/87006-02A; 249/87011-02A)
- b.
The Core Spray System piping analysis performed by Nutech Engineers inaccurately specified the stress intensification factor for a tee, modelled a new restraint on the pipe, but never designed or installed this restraint and incorrectly specified schedule 30 instead of schedule 80 piping. These errors contributed to the pipe exceeding the design stress limits.
(237/87006-28; 249/87011-28)
This is a Severity Level IV violation (Supplement 1).
Pursuant to the* provisions of 10 CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each violation: (1) corrective action taken and the results achieved; (2) corrective action to be taken to avoid further violations; and (3) the date when full compliance will be achieved.
Consideration may be given to extending your response time for good cause shown.
Dated J. J. Harrison, Chief Engineering Branch