ML20040F558
| ML20040F558 | |
| Person / Time | |
|---|---|
| Issue date: | 01/18/1982 |
| From: | Fox D, Hale C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20040F550 | List: |
| References | |
| REF-QA-99900525 NUDOCS 8202090343 | |
| Download: ML20040F558 (14) | |
Text
.
ORGANIZATION'-
READING, PENNSYLVANIA REPORT NO.: 99900525/81-03 INSPECTION DATE(S): 10/13-16/81 INSPECTION ON-SITE HOURS: 35 CORRESPONDENCE ADDRESS:
Gilbert / Commonwealth Utilities Group ATTN:
Mr. H. Lorenz, Group Vice President Post Office Box 1498 Reading, PA 19603 ORGANIZATION CONTACT:
Mr. N. R. Barker, Vice President, QA TELEPHONE:
(215) 775-2600 PRINCIPAL PRODUCT: Architect Engineering and Consulting Services NUCLEAR INDUSTRY ACTIVITY: The total effort committed to domestic nuclear activities at the Reading facilities is approximately 34% of the 2700 person staff of the Utilities Group.
Major projects include Perry Units 1 and 2, V. C. Summer, Unit 1, and Three Mile Island, Unit 1 Restart.
.}_ M i/f/82 ASSIGNED INSPECTOR:
[T. F. Fox', React 6r Systems Section (RSS)
Date' OTHER INSPECTOR (S):
8
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APPROVED BY:
C. J. @ e, Chief, RSS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B; Topical Report GAI-TR-106; and the SARs for the Perry Nuclear Power Plant, Units 1 and 2, and the V. C.
l Summer Nuclear Station, Unit 1.
i B.
SCOPE:
Determine the status of committed corrective action and preventive measures for previous inspection findings and a special inspection to determine the cause, significance, and design corrective actions taken on:
(1) a South Carolina Electric and Gas Company 10 CFR Part 21 report that the V. C. Summer Nuclear Station, Unit 1, reactor building containment spray system sodium hydroxide liquid storage tank was not procured to l
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O'RGANIZADON: GILBERT / COMMONWEALTH, UTILITIES GROUP READING, PENNSYLVANIA REPORT NO.: 99900525/81-03 INSPECTION RESULTS:
PAGE 2 of 8 INSPECTION BASES AND SCOPE (cont'd) operate under vacuum conditions as required; (2) a Gilbert / Commonwealth 10 CFR Part 21 report that the anchor and barrier plates, which attach piping to the Perry Nuclear Power Plant, Units 1 and 2, ASME Code Class MC containment penetration sleeves, were not impact tested or normalized as required; and (3) a Tennessee Valley Authority 10 CFR Part 50.55(e) report that the snubber for seal bypass piping of a Sequoyah, Unit 2, reactor coolant pump was misoriented.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to G/C 10 CFR Part 21 report to NRC dated April 28, 1981, revision XI of specification SP-527-4549-00 does not specifically address the material requirements for the Perry, Units 1 and 2, anchor and barrier plates of contaihment mechanical piping penetration assemblies as committed.
2.
Contrary to section 3.0 of ANSI N45.2.11, design specifications 597-044461-00, 592-044461-00 and 527-4549-00 did not include all the required design input data (see sections E.1 and E.2 for details).
3.
Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50 and TVA procedures CEB-SQN-100 and SQN-0C-V-13.3, a piping systern analyst did not identify, nor note on the affected piping system stress isometric drawing 47K-406-128, that the spatial orientation of the X-axis used in stress calculation 0600154-08-17 was reversed from the conventional orientation.
Further, the checker failed to detect and record the error as required by the Sequoyah Nuclear Power Plant Checking Procedure.
4.
Contrary to Criterion V of Appendix 8 to 10 CFR Part 50 and section 4.1.4 of V. C. Summer procedure 4.2.1, Mechanical Department calculation 13.19 dated June 11, 1981, did not note the specific purpose or objective of the calculation nor the source of the negative three PSIA pressure differential and atmospheric pressure variation that were used as design inputs to the calculation.
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ORGANIZATION:
G P A
REPORT NO.: 99900525/81-03 INSPECTION RESULTS:
PAGE 3 of 8 C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (81-01):
Design errors and deficiencies were not detected by the design verification cycle nor during the performance of QA audits.
The report of the Task Force, established by G/C to investigate the effectiveness of the G/C design control process, was completed as committed.
The report contains 16 specific reco:uendations for improving the design control process and was submitted to the G/C QA Policy Committee for their consideration on October 16, 1981.
Similarily, design errors and deficiencies identified in this inpsection report were not detected by the design verification cycle nor during the performance of QA audits.
The effectiveness of the implementation of the Task Force's reco:nmendations for assuring that design documents are thoroughly reviewed and design verified continues to be a concern and will be further inspected during subsequent inspections.
2.
(Closed) Follow up (81-02):
Interdisciplinary review and design verification of certain electrical and I&C design drawings did not appear to be formalized in procedures as required by Sections 4.0 and 6.0 of ANSI N45.2.11.
G/C engineering management stated that Design Control Procedure DCP:1.25 and Instrument and Control Section Procedure DP:0423-2.0.0 will be revised by October 25, 1981, to require timely interface review and design verification of electrical and I&C design drawings.
3.
(Closed) Follow up (81-02):
Design documents may not have been verified in a timely manner.
This item resulted in nonconformances B.2 and B.4 above.
4.
(Closed) Follow up (81-02):
The QA program for design corrective action may not totally satisfy Section 9.0 of ANSI-N45.2.11-1974.
ORGAN 6 M GlLBERT/ COMMONWEALTH, UTILITIES GROUP READING, PENNSYLVANIA REPORT NO.: 99900525/81-03 lNSPECTION RESULTS:
PAGE 4 of g G/C engineering management stated that a new DCP would be issued by December 31. 1981, to implement all of the requirements for design corrective action delineated in Section 9.0 of ANSI N45.2.11-1974.
G/C QA management stated that an annual assessment of the congruency of G/C design control procedures with committed ANSI standards would be conducted and documented.
E.
OTHER FINDINGS OR COMMENTS:
1.
Special inspection to determine the cause, significance, and design corrective action taken by G/C on a South Carolina Electric and Gas Company 10 CFR Part 21 report that the V. C. Summer Nuclear Station, Unit 1, reactor building containment spray system sodium hydroxide tank was not procured to operate under vacuum conditions as required.
Review of svailable documentation and interviews with cognizant engineering and management personnel indicate that the installed V. C. Summer, Unit 1, containment spray sodium hydroxide storage tank was not designed for vacuum operation, the tank's vacuum relief valves were not specified for use undar outdoor environmental conditions, nor were the valves required,to operate after a design basis seismic event. The safety concern was that when the containment spray system is activated, sufficient sodium hydroxide to remove the iodine that may be present in the containment's atmosphere may not be available due to failure of the vacuum relief valves to allow drawdown of the liquid from the tank.
The tank was ordered on December 9, 1975, under G/C specification SP-597-044461-00 to operate in the pressure range from 0 PSIG to
+ 3 PSIG under the provisions of Subsection ND of the ASME Code.
Paragraph ND-3821.1 permits a maximum negative pressure differential (vacuum) of 0.5 ounce per square inch unless the tank was specifically designed otherwise.
The vacuum relief valves were ordered on September 9, 1976, under G/C Specification 592-044461-00. The specification reouired the valves to operate at a negative pressure of 0.5 PSIA (not 0.5 ounce per square inch) and omitted the requirements for operability of the valves when installed in an outdoor environment and for operability after being subjected to a design basis seismic event as required by Regulatory Guide 1.48.
Seismic qualification (of the pressure boundary) was specified.
The tank and associated vacuum relief valves were installed outside the Auxiliary Building in accordance with G/C design drawings.
Startup Field Report SFR-3843 identified that dirt, water, and rust were observed on the seat of the valves.
Subsequent investigation by G/C engineering revealed the design deficiencies discussed above.
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GILBERT / COMMONWEALTH, UTILITIES GROUP ORGANIZATION.'
READING, PENNSYLVANIA REPORT NO.:
99900525/81-03 INSPECTION RESULTS:
PAGE 5 of 8 G/C engineering determined by analysis (Mechanical Department Calculation 13.19 dated June 11, 1981) that the as-built tank could withstand a negative three PSIA pressure differential and revised specification 597-044461-00 on April 21, 1981. The inspector could not determine if the manufacturer of the tank was advised of this change.to the specification. The inspector noted that the revised specification could permit operation of the tank outside of the limits of:
(1) the design pressure as specified to the fabricator of the tank and the design pressure limitations defined in the ASME Code (ND-3821.1); the pressures certified in the existing manufacturer's Data Report (ND-3812); and (3) the "N" stamping of the tank itself (NA-8000).
Valve Specification 592-044461-00 was revised on April 27, 1981, to include the requirement for operability after a design basis seismic event.
The valves themselves have reportedly been returned to the manufacturer for qualification testing and their status could not be determined during the inspection.
The specification was not revised to include a requirement, for operability in an outdocr environment since G/C engineering determined that the valves (but not the tank) should be relocated to inside the Auxiliary Building and revised the piping and structural drawings accordingly.
Two nonconformances were identified in ths area of the inspection (see B.2 and B.4 above).
2.
Special inspection to determine the cause, significance, and design corrective action taken by G/C on a G/C 10 CFR Part 21 report that the anchor and barrier plates, which attach piping to the Perry Nuclear Power Plants, Units 1 and 2, ASME Code Class MC containment penetration sleeves, were not impact tested or normalized as required.
Review of available documentation and interviews with cognizant personnel indicate that Perry, Units 1 and 2, containment piping penetration anchor and barrier plates were not designed in accordance with SAR commitments nor were they fabricated (some of which had been installed) in accordance with applicable ASME Code requirements for components to be subsequently welded to Code Class MC metal containment components.
The safety concern was that the weldment and/or the plates could fail (" brittle fracture") under postulated accident conditions and result in potential offsite exposures exceeding allowable limits.
~
GlLBERT/ COMMONWEALTH, UTILITIES GROUP ORGANNON:
READING, PENNSYLVANIA REPORT NO.:
99900525/81-03 INSPECDON RESULTS:
PAGE 6 of 8 The anchor and barrier plates were ordered as integral parts of certain piping spool pieces on July 4,1977, under G/C Specification SP-527-4549-00.
The specification did not specifically require impact testing as required by paragiaph NE-4431 of the ASME Code for Safety Class 2 and 3 piping and component support materials that are permanently attached by welding to Code Class MC components.
G/C Specification SP-527-4549-00 requires that impact tested weld material be used for all ASME Class MC work.
The installation contractor questioned the necessity for its use to weld the anchor and barrier plates to the containment Class MC penetration sleeves since the plates themselves were not impact tested.
Subsequent investigation by G/C engineering revealed that not only was impact testing (or normalization) not imposed by the specification, but that 4 of the 11 heats used to fabricate the plates yielded unsatisfactory results when impact testing was performed.
The affected spool piece assemblies were returned to the fabricator for repair or replacement with plate material from acceptable heats.
Specification SP-527-4549-00 was revised (Revision XI) to include a requirement that " materials which are permanently attached by welding to Class MC components shall meet the requirements of Section NE 4431 of the Code." However, the specification did not identify:
(1) which materials were affected; (2) which materials were permanently attached by welding to Class MC components; and (3) that the containment penetration sleeves (supplied by a different manufacturer) were Class MC components.
G/C engineering stated that the specification will be revised (Revision XII) to specifically require the anchor and barrier plates to be furnished in accordance with paragraph NE-4431 of the Code.
The related G/C design drawing 04-4549-8-312-633 will be revised (revision F) to clearly identify the Code Cla s of all affected components and their interfaces (weldments).
The inspector noted that the current design specification and the as-built condition of the penetration assemblies are in contrast with paragraphs 3.8.2.1.6.a.8 and 3.8.I. 1.6.a.7 of the Perry PSAR and FSAR (respectively) which state that " Mechanical penetration primary barriers are designed, fabricated, tested and inspected in accordance with ASME Code,Section III, Class MC (Subsection NE)."
Two nonconformances were identified in this area of the inspection (see B.1 and B.2 above).
ORGANIZATION: GILBERT / COMMONWEALTH, UTILITIES GROUP READING, PENNSYLVANIA REPORT NO.:99900525/81-03 INSPECTION RESULTS:
PAGE 7 of 8 3.
Special inspection to determine the cause, significance, and design corrective action taken by G/C on a Tennessee Valley Authority 10 CFR Part 50.55(e) report that the snubber for seal bypass piping of a Sequoyah, Unit 2, reactor coolant pump was misoriented.
Review of available documentation and interviews with cognizant personnel indicate that as a result of a design error, the snubber for the 3/4 inch bypass piping for the number one seal of reactor coolant pump number three was installed in a direction to retard pipe motion opposite to that observed during hot functional testing.
The safety concern was that pipe motion in a direction opposite to that for which the snubber was oriented could exceed the working range of the snubber and significantly increase the stress in the piping and could cause the pipe to break.
The snubber at 08-17-R0 node 21A on TVA isometric drawing 47K-406-128 and hanger number 2-CVTH-934 were designed as a result of stress calculation / piping analysis TVA-0600154-08-17 performed by G/C on December 8, 1980. The analysis was requested by TVA to update an earlier EDS Nuclear analysis to include the effects of newly specified seismic, thermal, and containment radial pulse pressure movement, displacement, and acceleration histograms.
The piping stress analyst apparently reversed the positive direction of the X-axis of the local coordinate system, from the conventional direction, to simplify the analysis at pipe location 08-17-R0 node 21A.
The analyst did not note the change on the system stress isometric drawing as required by TVA procedures.
Further, the checker failed to detect and record the error during the required check of the system design analysis.
G/C corrected the calculation on May 1, 1981, revised drawing 47K-406-128, and transmitted both to TVA in May 1981.
None of the TVA procedures nor calculation packages were available at the Reading Office for examination by the inspector.
However, a documented review by G/C personnel on October 16, 1981, of approximately twenty percent of the calculations performed by the G/C Oak Ridge Engineering Office, did not reveal any additional examples of this type of error.
G LBERT/ COMMONWEALTH, UTILZTIES GROUP ORGANIZATION *-
READING, PENNSYLVANIA RErORT NO.: 99900525/81-03 INSPECTION RESULTS:
PAGE g gf g Examination of design verification records for V. C. Summer piping stress analyses / calculations SCY-112nd CS-18 that were performed and checked at the Reading Office revealed no further examples of this type of error.
However, the checker did identify 11 items in SCY-ll and 3 in CS-18 that required resolution prior to issue of the calculations.
The undetected error in TVA calculation 0600154-08-17 appeared to be an isolated event.
One nonconformance was identified in this area of the inspection (see B.3 above).
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