ML20155G553
| ML20155G553 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/02/1986 |
| From: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20155G550 | List: |
| References | |
| EA-86-009, EA-86-9, NUDOCS 8605060212 | |
| Download: ML20155G553 (14) | |
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APPENDIX A NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES Texas Utilities Electric Company Docket Nos. 50-445 Comanche Peak Steam Electric Station 50-446 Units 1 and 2 Permit Nos. CPPR-126, CPPR-127 EA 86-09 During an NRC Technical Review Team (TRT) Inspection which began on July 9, 1984, violations of NRC requirements were identified.
In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1985) the Nuclear Regulatory Commission proposes to impose civil penalties pursuant to section 234 of the Atomic Energy Act of 1954, as amended.
("Act"), 42 U.S.C. 2282, PL 96-295, and 10 CFR 2.205. The particular violations and associated civil penalties are set forth below:
I.
Violations Assessed a Civil Penalty A.
10 CFR Part 50, Appendix B, Criterion II requires, in part, that the quality assurance program provide control 'over activities affecting the quality of the identified structures, systems, and components to,
an extent consistent with their importance to safety.
It further requires that this program provide for indoctrination and training of personnel performing these activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained.
The Comanche Peak Steam Electric Station (CPSES) Final Safety Analysis Report (FSAR), by Amendment 15 (April 30,1981) commits to NRC Regulatory Guide 1.58, Revision 1, " Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel," with minor modifications. This Regulatory Guide endorses, with comments, ANSI N45.2.6-1978, " Qualification of Inspection, Examination, and Testing Personnel for Nuclear Power Plants." ANSI N45.2.6-1978 provides guidelines and criteria for the evaluation and qualification of inspection personnel.
1.
ANSI N45.2.6-1978 requires in paragraph 2.4 that the ~ qualification of personnel be certified in writing in an appropriate form for the basis used for certification of qualification, including education and employment experience.
Paragraph 3 of this standard defines the minimum capabilities that qualify personnel to perform inspections, examinations, and tests.
Contrary to the above, since April 30, 1981, the quality assurance program did not adequately ensure that quality assurance / quality control QA/QC inspectors were appropriately qualified and trained to inspect activities affecting quality. Of 102 ASME and non-ASME inspector qualification records reviewed by the TRT, twenty percent did not contain verification of education or employment experience to substantiate their qualifications as required.
In addition, N
N G
Notice of Violation the TRT identified seven instances where the inspectors did not meet the minimum capabilities of the qualification requirements defined in ANSI N45.2.6-1978. These individuals were certified to the Level II capability within one to eight months of their transfer to the quality control program, even though they j.
possessed no prior inspection experience.
The recommended years of related experience defined in the standard for the Level II capability for these individuals was three years.
Reference (Ref): Safety Evaluation Report, Supplement (SSER) 11, 0-107.
2.
ANSI N45.2.6-1978 requires that quality assurance program personnel who plan and set up inspections, supervise or maintain surveillance over inspections, supervise and certify lower level personnel, report inspection results, and evaluate the validity and acceptability of inspections be certified to Level II capability.
Contrary to the above, at the time of the TRT inspection, the coating quality assurance program personnel who planned and set up inspections, supervised or maintained surveillance over inspections, supervised and certified lower level personnel, reported inspection results, and evaluated the validity and acceptability of inspections were certified to a Level I capability in lieu of the required Level II.
Ref:
SSER 9, M-121.
B.
10 CFR Part 50, Appendix B, Criterion III requires in part that measures must be established to assure that applicable regulatory requirements and design bases, as defined in 650.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies, are correctly translated into specifications, drawings, procedures, and instructions. These measures must include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. The design control measures must provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
In addition, design changes, including field changes, must be subject to design control measures commensurate with those applied to the original design.
CPSES FSAR, Section IA(b) commits to the guidance provided in Regulatory Guide 1.64, " Quality Assurance Requirements for the Design of Nuclear Power Plants," which endorses ANSI N45.2.11 (Draft 2, Revision 2-5/73).
1.
CPSES FSAR Section 8.3.1.4 specifies that the criteria used to establish the minimum requirements for preserving the independence of redundant Class IE systems are defined in IEEE Standard 384-1974,
" Trial-Use Standard Criteria for Separation of Class IE Equipment and Circuits."
I Notice of Violation IEEE Standard 384-1974, Section 5.6.2 requires in part that the minimum separation distance between redundant Class IE equipment and wiring internal to the control switchboards be established by analysis of the proposed installation. Where the control switchboard materials are flame retardant and an analysis is not performed, the minimum separation distance shall be six inches.
In the event the above separation distances are not maintained, barriers must be installed between the redundant Class IE equipment and wiring.
Contrary to the above, at the time of the TRT inspection, the applicant had failed to satisfy the minimum separation requirements of IEEE Standard 384-1974.
The TRT inspection identified several instances where these requirements were not translated into instructions for separation in the Unit I control room.
Both safety and nonsafety-related cables were in direct contact with other safety-related cables within flexible conduits and no analysis was provided that demonstrated the acceptability of the design and installation.
Ref: SSER 7, J-37.
2.
ANSI N45.2.11 " Quality Assurance Requirements for the Design of Nuclear Power Plants," in paragraph 8, requires that design changes be justified and subjected to design control measures commensurate with those applied to the original design.
Contrary to the above, in July 1983, the applicant reported to the NRC that during hot functional testing excessive air temperatures were documented near the vessel flange and in the ex-core detector wells. A subsequent review by the TRT indicated that prior to the installation of the reactor pressure vessel reflective insulation, the vendor requested and incorporated a design change to permit the installation of the reflective insulation support channel outside the insulation.
This design change effectively reduced the gap between the vessel insulation assemblies and the shield wall thus restricting the cooling air flow. This design change was not justified and subjected to design control measures commensurate with those applied to the original design in that had this condition not been detected during start-up testing, the integrity of the reactor vessel i
shield wall could not be assured after long-term exposure to elevated temperatures.
In addition, the sensitivity of the source range detectors in the neutron detector well area could not be assured.
Ref: SSER 8, K-99.
3.
CPSES FSAR Section 3.78.2.8 requires that non-Category I equipment and components located in Seismic Category I buildings are to be investigated by analysis or testing (or both) to ensure that l
structural integrity is maintained under the prescribed earthquake loading so that earthquakes do not adversely affect the integrity i
or operability (or both) of any designated Seismic Category I structure, equipment, or component.
t
Notice of Violation Contrary to the above, the TRT inspection identified non-seismic components in the Unit I control room ceiling and other Category 1 areas that were not analyzed in accordance with FSAR Section 3.78.2.8 that ensures earthquakes do not adversely affect the integrity or operability (or both) of any Seismic Category I structures, equipments, or components in these areas.
In addition, the analyses performed for the Category II light fixtures, the non-seismic drywall ceiling, and the lack of analysis for the non-safety-related conduits two inches (or less) in diameter, did not ensure that the structural integrity is maintained under the prescribed earthquake loadings.
Ref: SSER 8, K-83.
4.
CPSES FSAR, Sections 3.7B.2.8 and 3.78.3.13 require that the effects on seismic Category I piping from non-Category I piping and structures be considered.
Contrary to the above, the NRC Region II Special Review Team (SRT) and the TRT identified a Category I and non-Category I interaction for which an analysis could not be produced that showed compliance with the CPSES FSAR requirements for the piping at the Electrica.1 Control Building / Turbine Building interface.
Ref: SSER 10, N-238.
5.
CPSES FSAR, Section 3.78.2.8 requires that non-Category I equipment and components located in seismic Category I buildings are investigated by analysis or testing, or both, to ensure that under the prescribed earthquake loading, structural integrity is maintained, or the non-Category I equipment and components do not adversely affect the integrity or operability, or both, of any designated seismic Category I structure, equipment, or component to the extent that these seismic Category I items cannot perform their required functions.
Contrary to the above, the TRT found that the design analysis for non-Category I equipment effects on seismic Category I structures, equipments, or components was incomplete. The support installation for nonsafety-related conduits less than or equal to 2 inches was inconsistent with seismic requirements and no evidence could be found that substantiated the adequacy of the installation for nonsafety-related conduit of any size.
Ref: SSER 7.
6.
ASME Section III, NF-4725 requires that threaded fasteners, except high-strength bolts, be provided with locking devices to prevent loosening during service.
Brown & Root Instruction QI-QAP-11.1-28, i
Revision 25 in Section 3.7.1 requires that exposed threads be free of extraneous material.
i l
I Notice of Violation Contrary to the above, measures were not established to ensure that the standards for locking devices were specified and included in design documents. On May 24, 1984 Texas Utilities Electric Company (TUEC) engineering issued a memorandum (CPPA 38997) that approved paint, when applied to Unit 1 component supports, including fasteners, and when set or hardened, would act on bolt and nut threads to prevent the nut from loosening.
In addition, suitability testing did not justify the use of paint as a substitute locking device per the ASME code. The use of paint in this manner is contrary to the ASME and site procedural requirements.
Ref: SSER 11, 0-244.
7.
ANSI N45.2.11, in paragraph 4.1 requires that design activities be prescribed and accomplished in accordance with procedures of a type sufficient to assure that applicable design inputs are correctly translated into specifications or procedures. TUEC Procedure CP-EP-4.0, " Design Control," Revision 3 dated July 11, 1982 requires that design inputs, on which final design is based, be identified, documented, and approved.
Contrary to the above, at the time of the TRT inspection, engineering criteria defined in Bechtel Corporation Specification 10466-M-204, Appendix D governing the cold springing of piping systems during installation was used at CPSES as the basis for the final design of the piping systems.
However, this criteria was never formally identified, documented, or authorized in CPSES TUEC engineering documents.
Ref:
SSER 10, N-99.
8.
ANSI N45.2.11 in paragraph 3.1 requires that design inputs, such as design bases, be identified, documented, and their selection reviewed and approved.
Changes from specified design inputs, including the reasons for the changes, shall be identified, approved, documented and controlled.
Contrary to the above, at the time of the TRT inspection, the applicant failed to adequately identify design bases and inputs including specific Design Basis Accident (DBA) test reports, and failed to properly perform and document review and analysis of design and design inputs, especially design changes.
For example, allowable coating thicknesses applied to the inside of the Containment Liner were repeatedly changed without engineering evaluation and review to demonstrate that the coatings of different thickness would survive testing under DBA conditions.
Ref:
SSER 9, M-11.
Notice of Violation i 9.
CPSES FSAR, paragraph 6.18.2, states, in part, " Coating systems i
used on exposed surfaces in the Containment have been qualified in accordance with ANSI N101.2 and N512 and are applied in accordance with NRC Regulatory Guide 1.54."
The CPSES FSAR, Section 17A, including Table 17A-1, and the CPSES Specification 2323-AS-31, in paragraphs 2.0b and 3.0a, require that coatings shall be tested j
and approved for application in areas exposed to radiation by Oak Ridge National Laboratories and the coating manufacturer in accordance with ANSI N101.2 and N512.
Contrary to the above, at the time of the TRT inspection, the 1
applicant could not provide evidence to demon _ strate that coating systems used at CPSES had been tested and qualified by Oak Ridge National Laboratories and the coating manufacturer, in accordance with ANSI N101.2 and N512.
This deficiency applies to the originally specified coating systems and to subsequent design changes.
Ref:
SSER 9, M-50.
C.
10 CFR Part 50, Appendix B, Criterion X requires, in part, that a program for inspection of activities affecting quality be established and executed by or for the organization performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity.
1.
Texas Utilities Generating Company (TUGCO) Procedure QI-QP-11.3-28 Rev. 21 " Class IE Cable Terminations," paragraph 3.1.c states that "All Class IE and associated cable splices and terminations j
that utilize splice connectors shall be witnessed."
{
Contrary to the above, the TRT reviewed twelve quality control inspection reports for butt splices of 600 volt control and instrument connections and found three 'ncidents where the applicant's quality control inspector failed to witness the j
splice as required in paragraph 3.1 of procedure QI-QP-11.3-28.
Ref: SSER 7, J-29.
2.
Gibbs & Hill Specification 2323-MS-46A, Revision 5 requires that nuclear safety-related Class 1, 2, and 3 pipe hangers and supports meet the requirements of the ASME Code, 1974 Edition,Section III l
Subarticle NF-431 which states that " Defects in materials which were. accepted on delivery or which are discovered during the process of fabrication or installation may be eliminated or i
repaired by welding, provided the defects are removed, repaired, and examined in accordance with the requirements of NF-2500 for the applicable product form." Subarticle NF-2500 requires that defects may be repaired as permitted by the material specifications.
i Brown & Root Procedure WES-029, " Welding Specification for Field j
Fabrication and Erection of Structural Steel" delineates the i
instructions and documentation required to perform weld repairs j
(i.e., plug wells of misdrilled holes).
)
Notice of Violation Contrary to the above, cable tray supports were found by NRC Region IV staff personnel to contain undocumented plug welds.
It was not possible to determine if the plug weld repairs were acceptable because quality control inspections were not performed to ensure that the welds were performed in accordance with the applicable codes and procedures.
Ref: SSER 10, N-57.
3.
TUEC procedure QI-0P-11.14-1, " Inspection of Site Fabrication and Installation of Structural and Miscellaneous Steel," delineates inspection criteria for support bolts. Work package MRB-0550-013-RB authorized the cutting of 1-1/2 inches off the 9-inch length of 144 bolts.
The bolts were purchased to hold the steam generator lateral supports to the wall plates but were ordered 1-1/2 inches too long.
Contrary to the above, required quality control inspections for the installation of the steam generator upper lateral restraint anchor bolts were not performed.
Ref: SSER 10, N-57.
4.
TUGC0 Procedure QI-QP-11.10-1, Revision 29, Paragraph 3.5.2 includes the requirement to inspect a support for configuration.
TUGC0 Procedure QI-QP-11.21-1, Revision 8, " Requirements of Visual Weld Inspection," sets forth the criteria and requirements to be used when performing visual inspections of welds and other applicable instructions for conduit support and cable tray hanger inspections.
Contrary to the above, the TRT performed an independent inspection of previously accepted welds and identified the following deficiencies that indicated that quality control inspections in this area were inadequate as evidenced by the following:
(a) The TRT found that cable tray hanger CTH 5824 (Containment Building) had been fabricated with forty more stiffeners and eighty more welds than required or shown on drawing FSE-00159, sheet 5824, 2 of 2, Detail L Inspection Report ME-1-0006155 verifiedfinalQCinspecti$n. and acceptance on January 3, 1984.
(b) Cable tray hanger CTH-6742 (Auxiliary Building), Clip MK-12, should be 6 inch x 6 inch x 3/4 inch angle stock in accordance with FSE-00159, sheet 6742.
The actual "as-built" flange thickness of Clip MK-12 was 3/8 inch.
(c) During inspection of hanger CTH-6742, the TRT found that two structural welds were made in the wrong location. The 3/16 i
inch shop welds which join MK-10 and MK-11 were made horizontally instead of vertically, as shown on drawing FSE-00159, sheet 6742.
Ref: SSER 11, 0-244.
a Notice of Violation 5.
Brown & Root Procedure QI-QAP-11.1-28, Revision 19. " Fabrication, Installation, and Inspection of Safety Class Component Supports" provides instructions for material dimensional control and fabrication tolerances.
The procedure limits base plate hole centerline locations to 11/4 inch or as shown on the design drawing.
Contrary to the above, quality control inspections were inadequate in that the TRT discovered at the time of its inspection that the horizontal member of support CC-1-126-010-F33R was three inches lower at its centerline relative to the upper bolthole centerline than shown on the vendor-certified drawing.
Other supports with similar hole-location violations included CC-X-039-007-F43R, CC-1-126-011-F33R, and CC-1-126-012-F33R.
Ref: SSER 11, 0-244.
6.a. Brown & Root Procedure QI-QAP 11.1-28, Revision 25, " Fabrication, Installation and Inspection of Safety Class Component Supports," paragraph 3.7.3.1 states that "a sufficient number of spacers shall be used to prevent the spherical bearings from becoming dislodged..." and "in no case shall the resulting gap be more than the thickness of one vendor-supplied spacer."
Contrary to the above, quality control inspections were inadequate in that the TRT inspection identified an excessive free gap between spherical bearing and washers on the sway strut assembly of support CC-1-126-015-F43R. Other supports with similar bearing gap anomalies identified during the TRT inspections included supports RC-1-052-016-C41K, RC-1-052-020-C41K, and MS-1-416-001-533R.
- b. QI-QAP-11.1-28, Revision 22, paragraph 6.1 states in part that
"... bearing internal and external surfaces shall be free of rust and foreign material, and bearings shall move freely within the housing."
Contrary to the above, quality control inspections were inadequate in that the TRT inspection identified paint contamination in the bearings of both snubber assemblies on Class 1 component support SI-1-090-006-C41K of the Unit 1 Safety Injection System that severely obstructed the bearing cavities and limited their movement.
A similar condition existed on support MS-1-416-002-533R.
Ref: SSER 11, 0-244.
7.
TUGC0 Instruction CP-EI-4.5-1, Revision 10 " General Program for As-built Piping Verification" requires verification in the field to ensure that actual hanger mark numbers agree with the mark numbers shown on the drawing and that the hanger type agrees with that shown on the support drawing.
k 1
Brown & Root Procedure QI-QAP-11.1-28, Revisions 19 and 24, require that at installation inspection, the quality control inspector verify the hanger number, material type, grade and heat number, and that vendor-supplied NPT-stamped component supports bear markings traceable to the design drawing.
I Contrary to the above, the TRT inspection identified in six instances (from an inspection of 42 supports) where these procedural requirements were not followed during the final quality control inspections. These instances are as follows:
(a) Model numbers of installed snubbers for pipe support SI-1-090-006-C41K did not match the model number on the design drawing. A similar problem existed with pipe support RC-052-020-C41R.
(b) A replacement part (sway strut eyerod) for pipe support CT-1-013-014-532R had no apparent material identification either on the hardware or in the documentation package for 4
1 the support. The Material Identification Log did not list any identification traceable to the origin of the replacement part. A similar problem existed with pipe supports CC-1-126-012-F33R, CC-X-039-005-F43R, and AF-1-035-011-533R.
Ref:
SSER 11, 0-244.
J 8.
Brown & Root procedure QI-QAP-11.1-28, Revision 25 defines criteria for the examination of welds, including inspection parameters for acceptable weld sizes. ASME B&PVC Code Section III, Subsection NF. Subarticles NF-4424 and NF-5360 set forth acceptance standards for the examination of welds.
3 Contrary to the above, the following deficiencies found during the TRT inspection that were not identified during the applicant's quality control inspection process for the following piping supports:
(a) Support AF-1-001-001-533R had porosity, insufficient weld leg, incomplete welds, and insufficient fill.
(b) Support CT-1-013-014-532R exhibited excessive overgrinding of welds which resulted in notching of the sway strut rear brackets.
(c) Support AF-1-002-702-S33R had two more welds than required.
The extraneous welding was not documented on the "as-built" drawing. One of the required welds, was undercut (1/16 inch - 3/32 inch deep, for a length of 2 inches) beyond limits of acceptance.
I h
6 Notice of Violation (d) Support Drawing CCl-126-013-F33R required a 1/2 inch fillet weld to connect item 5 to item 6.
This weld was omitted in the field.
(e) Support CC-1-126-013-F33R had some welds performed with no quality control inspector initials or signature on the corresponding blocks of the weld data card for the support inspection package.
(f) Support CC-X-039-007-F43R, had a 5/16-inch all-around fillet weld with an approximately 1/16-inch undersized weld leg across the top.
(g) Support RH-1-006-012-C42R had an all-around 1/4-inch fillet weld connecting item 5 to item 7 which was undersized by 1/32 inch to 1/16 inch across the top.
(h) Support AF-1-037-002-533R exhibited a 1/16-inch to 3/32-inch reduction in plate thickness and weld size due to excessive grinding of the weld at the base plate. Base material thickness of the support plate was reduced beyond the limits of acceptance in three locations.
Ref: SSER 11, 0-244.
D.
10 CFR Part 50, Appendix B, Criterion XVI requires, in part, that measures be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
In the case of significant conditions adverse to quality, the measures must assure that the cause of the condition is determined and corrective action taken to preclude repetition.
1.
Brown & Root Procedure QI-CP-QCI-2.4-9 delineates the inspection and documentation requirements to ensure the removal of elastic joint filler (rotofoam) material between seismic Category I structures.
In addition, procedure QI-QP-11.0-3 details the inspection activities for Seismic Air Gaps.
Contrary to the above, measures were inadequate to assure that conditions adverse to quality were promptly identified and corrected in that quality control inspections between September 14, 1978 and October 17, 1978 documented that five of six seismic Category I structure gaps contained foreign material.
These unsatisfactory conditions were not officially resolved until April 18, 1983 in response to Nonconformance Report C-83-01067.
However, during the TRT inspection, foreign material was again i
3 identified between the Unit I safeguards building and the j
auxiliary building. The continued existence of foreign material l
in the air gaps indicates that measures were ineffective in assuring that nonconformances were promptly identified and corrected.
Ref: SSER 8, K-75.
Brown & Root Procedure CP-QAP-16.1, " Control of Nonconforming Items," requires that the approval of the disposition of Nonconformance Reports (NCR's) be reviewed for adequacy and conformance to applicable specifications, procedures, and code l
requirements.
Contrary to the above, the dispositions of NCRs M-4015S, Revisions 0-5 and NCR M-4942S were inadequate.
Their dispositions did not address fit-up or ASME code requirements nor the stress effects resulting from the out-of-roundness which occurred during installation. These NCRs are related to the installation of a 10-inch Spool Piece CT-1-SB-014, piece number 38, installed in the Unit 1 Containment Spray System CT-1-012-301R-2 in September 1982. At the time of installation, the item was on hold per NCR M-4015S because the pipe was 1/2 inch out-of-round and in excess of the 1/8 inch allowed by the ASME Code for 10-inch pipe. After the installation, NCR M-49425 and NCR M-40155 Revision 5 dispositioned the pipe "use-as-is" based on acceptable fit-up inspection results.
Ref:
SSER 10, N-155.
3.
Brown & Root Procedure CP-QAP-12.1, " Inspection Criteria and Documentation Requirements prior to N-5 System Certification,"
paragraph 3.8 and attachment 5 required reinspection of skewed welds for size determination during the inspection of the hangers.
Brown & Root Procedure QI-QAP-11.1-28, paragraph 3.5.5.5 and attachment 23 provide requirements and techniques for the proper inspection methods of welds that exhibit a skewed profile.
Contrary to the above, on December 28, 1983, TUEC failed to provide adequate corrective action in their reinspection program i
for ASME Class 1, 2, and 3 component support skewed welds.
Certain types of skewed welds, those existing at the intersection of curved members used as structural members, were not included l
in the skewed weld reinspection program for component supports.
l Ref:
SSER 10, N-202.
i l
Collectively, Violations A-D have been categorized as a Severity Level III problem (Supplement II).
(Civil Penalties - $200,000 assessed equally among the violations.)
II.
Violations Not Assessed a Civil Penalty A.
10 CFR Part 50, Appendix B, Criterion VI requires, in part, that l
measures be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality.
These measures assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel and are distributed to and used at the location where the prescribed activity is performed.
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Procedure DCP-3 "CPSES Document Control Program," Revision 18 states, in part, that "If, for any reason, a superseded document is retained, the face of the document must be stamped or marked
' VOID.' Wnen no longer required, superseded documents should be destroyed." DCP-3 also states, in part, that " Controlled documents affected by design change documentation shall be stamped as follows:
'THIS DOCUMENT AFFECTED BY DESIGN CHANGE.'
Contrary to the above, the following conditions were found in satellite 307 on July 31, 1984:
(a) Document package I-002-S-09 listed eighteen design changes outstanding; however, twenty changes were in the package, including one superseded and one voided package that were not appropriately marked.
(b) Two voided design changes were listed and included as current in design packages (CMC 62535, Revision 0 against MI-2C07 and DCA 13170, Revision 0 against MS-084) but the changes were not appropriately marked.
(c) Superseded drawing RH-1-SB-006, Revision 13 was found in the files on July 31, 1984, but was not appropriately marked.
(d) Drawings 2323-MI-2301-10, -2304-01, and -2304-05 were in the drawing satellite files on July 31, 1984, and Drawing D0-2-099-709-553R was in the hands of hanger craft personnel.
However, these drawings were not stamped "This document affected by design changes," even though each document was affected by changes.
Ref:
SSER 11, 0-51.
2.
The TUGC0 Quality Assurance program is included in Chapter 17 of the CPSES FSAR. Section 17.1.6, for document control, states that "TUGC0 has established requirements to assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel."
Contrary to the above, measures were not effective in assuring that drawings reflecting the as-built conditions were properly released for use in that the TRT identified six cables, five safety-related, that were not terminated in accordance with drawings effective at the time of the inspection.
Ref:
SSER 7, J-29.
B.
10 CFR Part 50, Appendix B, Criterion V requires in part that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstance and that these activities be accomplished in accordance with these instructions, procedures, or drawings.
1 Notice of Violation Brown & Root Procedure CP-CPM-9.9, "NNS Seismic Category II Supports,"
Section 3.1, states, in part, " Fabrication / Installation shall be accomplished in accordance with the Hanger Package.
The Hanger Package will consist of the BRH drawing and Weld Filler Material Log (WFML) for each hanger listed on the BRH."
Brown & Root Procedure CCP-21, Fabrication of Miscellaneous Steel, Section 3.1.3, requires that all work be accomplished in accordance with controlled drawings.
Contrary to the above, the TRT identified that iron fab shop work was s
performed to sketches and memos without the availability of the hanger package, traveler, or controlled drawing at the location where the activity was performed.
Ref:
SSER 11, 0-146.
Violations A and B have been categorized as a Severity Level IV problem (SupplementII).
Pursuant to the provisions of 10 CFR 2.201, Texas Utilities Electric Company is hereby required to submit to the Director, Office of Inspection and Enforcement.,
U.S. Nuclear Comission, Washington, D.C. 20555, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, within 60 days of the date of this Notice, a written statement or explanation in reply, including for each alleged violation:
(1) admission or denial of the violation; (2) the reasons for the violation if admitted; (3) the corrective steps which will be taken and the results achieved; (4) the corrective steps which have been taken to avoid further violations; and (5) the date when full compliance will be achieved.
If an adequate reply is not received within the time specified in this Notice, the Director, Office of Inspection and Enforcement, may issue an order to show cause why the license should not be modified, suspended or revoked or why such other action as may be proper should not be taken.
Consideration may be given to extending the response time for good cause shown.
Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath of affirmation.
Within the same time as provided for the response required above under 10 CFR 2.201, the Texas Utilities Electric Company may pay the civil penalties in the amount of Two Hundred Thousand Dollars ($200,000) or may protest imposition of the civil penalties in whole or in part by a written answer.
Should the Texas Utilities Electric Company fail to answer within the time specified, the Director, Office of Inspection and Enforcement, will issue an order imposing the civil penalties in the amount proposed above.
Should the Texas Utilities Electric Company elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalties, such answer may:
(1) deny the violation listed in this Notice in whole or in part; (2) demonstrate extenuating circumstances; (3) show error in this Notice, or (4) show other reasons why the penalties should not be imposed.
In addition to protesting the civil penalties in whole or in part, such answer may request mitigation of the penalties.
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Notice of Violation In requesting mitigation of the proposed penalties, the factors contained in Section V.B of 10 CFR Part 2, Appendix C (1985) should be addressed. Any written answer in accordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate by specific reference (e.g., citing page and paragraph numbers) to avoid repetition. The Texas Utilities Electric Company's attention is directed to the other provisions of 10 CFR 2.205 regarding the procedure for imposing a civil penalty.
Upon failure to pay any civil penalties due, which has been subsequently determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalties, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 234c of the Act, 42 U.S.C. 2282.
FOR THE NUCLEAR REGULATORY COMMISSION V
s M. Taylg [, Director a
fice of Inspection and Enforcement Dated a Bethesda, Maryland this ay of May 1986.