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- b' i JAN 2 o ym MEMORANDUM FOR: J. G. Davis, Director, Office of Nuclear Material Safety a; Safeguards FROM: Robbrt D. Martin, Regional Administrator
SUBJECT:
REFERENCE MATERIAL FOR CPRRG - NOTES OF H. PHILLIPS Enclosed is material prepared to support the review efforts of the CPRRG. The material provided consists of.the following:
i Notes or other materials prepared by H..Phillips relative to the technical issues about'which he had concerns and which were identified in attachment MM to the OIA r.eport.
. These materials have not been reviewed by Region IV management and are being provided to CPRRG as they were provided by Mr. Phillips.
Should you have any questions, do not hesitate to contact me.
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/W Robert D. Martin Regional Administrator cc w/o enclosures:
J. M. Taylor, IE H. Phillips i
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- N2 0 1987 I
MEMORANDUM FOR: J. G. Davis, Director, Office of Nuclear Material Safety & Safeguards ;
FROM: Robert D. Martin, Regional Administrator
SUBJECT:
REFERENCE M)YERIAL FOR CPRRG - NOTES OF Enclosed is material prepared to support the review efforts of the CPRRG. The teaterial previded consists of the following:
. Notes or other material prepared by M relative to the technical issues about which he had concerns and which were identified in attachment MM to the OIA report.
. These material have not been reviewed by Region IV mans ement and are being provided to CPRRG as they were provided by Should you have any questions, do not hesitate to contact me.
Robert D. Martin
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Regional Administrator cc w/o enclosures:
J. M. Taylor, IE P
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UNITED STATES
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$ f $11 RYAN PLAZA DRIVE, SUITE 1000 k% *[ ARLINGTON, TEXAS 70011 MEMORANDUM FOR: John Davis FROM: { p Reactor Inspector, Region IV
SUBJECT:
CONCERNS STATED IN REGION IV OIA REPORT As directed by Robert D. Martin, Region IV Regional Administrator, I have prepared an input to each technical issue identified by me and discussed in the OIA report. The issues, raised during'the April-June 1985 time period were based on the information discovered during the inspection and subsequent discussions with licensee personnel. Because of other assignments and priorities and since I do not have unlimited resources at my disposal, I have made no attempt to justify sqy position on the issues beyond that made when the inspection report was issued. The directed input is submitted as enclosure 1
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through 7.
My inspection philosophy is to establish inspection criteria based on 10 CFR 5'0 requirements, plant tech specs, SAR, and other licensee commitments, and industry codes. I consider codes imposed by 10 CFR 50 as law and do not believe that inspectors or supervisors have the authority to impose additional requirements or to refuse to enforce established requirements. This is not to say that management does not have authority to make decisions concerning regulatory matters, but if after discussions with the inspector, a difference in interpretation of a technical matter exists it is necessary to seek an official headquarters ruling so that enforcement will be uniform throughout industry.
j In testimony given by Westerman and Barnes, statements were made regarding my ability as an inspector because I would not agree with~their interpretation of certain ASME Code matters and with certain industry practices without l
headquarters concurrence. I have included as an attachment to this memo a brief summary of ray technical background, including statements made on a {
performance appraisal prepared by Barnes while he was my section chief in the j Vendor Program Branch. Of particular interest is the " exceeds standards" j rating'concerning a working knowledge of codes and regulations. l I
MY$$N Reactor Inspector
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Attachments (2) )
Enclosures (7) 1
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ATTACHMENT 1 Summary of Technical Background for
- BS in Machanical Engineering - 1958 - Texas Technological College
' GraduateStudyinMechanicalEngineering(15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />)-1961-1962- Wichita State University
- Aerospace Engineer - 1958-1971 - Propulsion background in aircraft and liquid fuel space vehicles l Aerospace and QA Engineer for D00 - 1971-D 82
- Extensive training and experience in QA
- Training and experience in contract administration, industry codes, specifications, and standards Taught D0D course in reading codes, standards, specifications, and drawings Planned and led complex audits of D00 contractors
- Received several awards, including outstanding performance appraisals and high quality pay increase NuclearEngineer(Inspector)inHRC-1982-Present Training and experience in a variety of technical areas, including ASME Code
- Performed inspections at a variety of vendors supplying products for nuclear power plants Performed a variety of inspections at nuclear plants under construction
- Planned and performed a variety of inspections at operating plants Led and performed EQ inspections at Region IV operating plants I
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SUMMARY
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ENCLOSURE 1 Inspection Report 85-07/05 Issue No. 1 (Attachment I to Attachment M-M) f i
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- 1. Definition of Issue
- "(a) Unit 2 reactor pressure vessel installation design criteria i recommended by the nuclear steam supply (NSSS). vendor, such as centering tolerances, levelness tolerances, and shoe to bracket clearances, were not {
included in installation specifications, procedures, and drawings; and (b) the criteria were specified in Construction Operation Traveler ME-79-248-5500, but were not treated as design engineering criteria as evidenced by an undocumented change of shoe to bracket clearance."
- 2. Understanding of Issue
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My understanding of this issue was based on information gathered at the time of the inspection and subsequent discussions with the license and his contre tors prior to the time of the 01A investigation. Inspection criteria and guidelines were based on the following:
- a. Regulatory Requirement - 10 CFR 50 Appendix B, Criterion III
" Measures shall be established to assure that applicable regulatory j requirements and the design basis . . . are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled . . . . Measures shall be established for the identification and control of design interfaces l and for coordination among participating design organizations. These measures shall include the establishment of procedures among partici-pating organizations for the review, approval, release, distribution and revision of documents involving design interfaces . . . . Design changes, including field changes shall be subjected to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization."
b..
Licensee Commitment - TUGCo QA Plan, Section 3, Revision 3, dated July 31, 1984, requires the appropriate site engineering organization tu prepare design specifications which are translated into instructions, procedures, drawings, or specifications.
- c. NRC Inspection Guidance - IE Module 50053-02.b(4) " Observe the installed reactor vessel and, based on a review of the installation !
i specifications, drawings, and work procedures, identify a number of requirements such as vessel support structures, vessel-to-support structure fittings, number and location of support structures and mounting pads, hold down devices, shimming devices, and alignment l '
requirement, and determine whether:
l
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work procedures have been or are being followed, '
installation is in accordance with speciff. cations and procedures, and inspection (QC) procedures are being followed."
At the time of this inspection, the RV was already in place, therefore, records of the placement were reviewed to verify proper placement. Construction and Operation Traveler No. ME-79-248-5500 which provided field instructions for installing the vessel was reviewed. Several drawings prepared by the site engineer and referenced by the traveler, were reviewed. None of the drawings included installation tolerances. A Westinghouse installation procedure referenced by the traveler was also reviewed.
The Westinghouse installation procedure contained certain installation ;
tolerances which were recommended, but it was obviously an uncontrolled document which had no provisions or requirements for change control. Certain tolerances (support bracket to shoe) l specified_on the traveler complied with NSSS. recommendations, but had been changed and established change control procedures had not been followed. It was my position that placement of the reactor vessel by the site construction contractor involved an interface between the site designed cavity and support structure and the NSSS contractor and that site specifications and drawings, prepared and controlled by ,
change control procedures, should have been developed to implement site engineering and NSSS contractor requirements. Any technical changes would then require a coordinated engineering change which would be implemented by the construction traveler.
- 3. Safety Significance This issue could result in a hardware problem because of the uncor. trolled tolerance change. It was impossible at the time of the inspection to make this determination and an analysis would be required by the licensee.
Equally as important is the generic implications of changing design or interface criteria with no documentation to support the change.
- 4. Understanding of How Issue Was Handled by Management This issue was documented as a violation by the inspector and changed to an unresolved item by management. Since an unresolved item requires no licensee response, it is not certain what will be involved in closing the item. Management stated to me that they considered the traveler to be an appropriate engineering document and provided no further instructions to me. Since I am no longer involved with CPSES, I am not aware of any NRC follow-up of this item.
l J
1 ENCLOSURE 2 Inspection Report 85-07/05 Issue No. 2 ( Attachment 1 to Attachment M-M) i
- 1. Definition of Issue
" Clearances between the reactor vessel and support bracket:; anf 9 pert f shoes were not within the tolerance stated in Construction Oper'+'on Traveler ME-79-248-5500 and the condition was not reported on a 4 Nonconformance Report."
- 2. T Understanding of Issue l
My understanding of this issue was based on information gathered at the '
time of the inspection and subsequent discussions with the licensee and his contractors prior to the time of the OIA investigation. Inspection criteria and guidelines were based on the following:
- a. Regulatory Requirement - 10 CFR 50 Appendix B, Criterion XV "Heasures shall be established to contro1 materials, parts, or components which do not conform to requirements in order to prevent Nonconforming items their inadvertent use or installation . . . .
shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures."
- b. Licensee Commitment (1) FSAR Section 17.1 requires written procedures to investigate nonconforming items, decisions on their disposition, and preparation of adequate reports.
1 (2) TUGCo QA Plan,Section V, Revision 2 dated May 21, 1981, implemented 10 CFR 50 Appendix B, Criterion XV.
(3) Brown and Root QA Manual, Section 16, dated March 27, 1985, requires that unsatisfactory conditions identified on process control documents are to be identified en a nonconforming report.
- c. NRC Inspection Guidelines - IE Module 50055-02.C(3)"Nonconformance Reports include status of corrective action or resolution."
This issue is closely related to issue number 1 and was identified during the same inspection. It was noted in concern number 1 that allowable tolerances between the reactor vessel support brackets and support shoes were changed without any apparent engineering justification. The as-left tolerance between support brackets and shoes exceeded both the original This condition had been accepted on the values and the revised values.
basis of a recommendation by an onsite NSSS contractor representative with
i 2 !
no documented engineering justification. Also, in violation of regulatory l
requirements and site procedures the noncompliance had not been documented and approved through the established site nonconformance system.
- 3. Safety Significance Without performing an analysis, the effect on. safety of this concern can only be surmised. It is presumed that the vessel support bracket to shoe tolerances were established by the HSSS contractor for a cold condition to aermit the vessel to grow in a uniform manner, without binding during teat-up. If tolerances are not correct or uniform between the different supports high localized stresses could occur and result in fatigue failure at the vessel nozzle over a period of time. I was concerned that tolerances had been exceeded'and that the condition had not been adequately documented and analyzed by engineering. The possible generic implication of other nonconformance remaining undocumented and unanalyzed is also a concern.
, Understanding of How Issue Was Handled by Management 4
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This issue was documented as a violation by me and changed to an unresolved item by management. Westerman stated that he did not believe that the issue was a violation and that given the time the NSSS contractor could probably verify that the as-left condition was acceptable. The issue was included in the report as unresolved pending documentation validating the final installation tolerances. No response was required of the licensee and.I received no instructions to follow-up. Since I am no longer involved with CPSES, I am not aware of any NRC follow-up of this item.
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i ENCLOSURE 3 i
Inspection Report 85-07/05 Issue No. 3 (Attachment 1 to Attachment M-M) l
- 1. Definition of Issue "There was no evidence that TUGCo had audited either Unit 2 reactor vessel installation specifications, placement procedures, actual hardware placement, or as-built records." '
l
- 2. ~ Understanding of Issue
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Ny understanding of this issue was based on information gathered at the l time of the inspection and subsequent discussions with the licensee and i hts contractors prior to the time of the OIA investigation. Inspection criteria and guidelines were based on the.following:
- a. Regulatory Requirement - 10 CFR 50 Appendix B, Criterion XVIII. ,
i "A comprehensive system of planned and periodic audits shall be I carried out to verify compliance with all aspects of the quality j assurance program and to determine the effectiveness of.the program." !
- b. Licensee Commitment - TUGCo QA plan, Section 18, Revision 2, dated July 31, 1984, implements 10 CFR 50 Appendix B, Criterion XVIII.
- c. NRC Inspection Guidance (1) IE Module 50051-02.022 " Determine whether the licensee has an established audit program (including plans, procedures, and schedule) covering safety-related work and control functions in the area of reactor and internals installation."
(2) IE Module 50055-02.e(1) " Records establish that the required audits were performed."
In conjunction with inspection of records for the reactor vessel, I made at least two attempts with different licensee personnel, who were NRC designated contacts, to determine whether QA audits of the reactor setting had been performed. This included either an audit of specifications or procedures, actual setting of the vessel, or of the installation records.
While it may be a close call to state, specifically that the reactor vessel setting must be audited Criterion XVIII requires that all aspects of the QA program be audited to determine its effectiveness and this appears to be supported by IE Module 50055.
This implies to me that all safety-related systems must be audited sometime during the construction process. This would appear especially important because of the involvement of more than one major contractor. A good QA program should be aware of problem areas and he conducted accordingly; however, in spite of at least two major reactor vessel 4
2 interface problems experienced in the past the licensee apparently failed to verify that the vessel had been properly installed for either unit 1 or unit 2. Typical problems experienced in the past were: (1)duringhot flow tests on unit 1 localized heating around the vessel occurred and modifications were required to improve cooling air circulation between the vessel and cavity wall and (2) prior to setting unit 2 vessel it was 1 discovered that the vessel mounts were misplaced by 45 degrees in the cavity walls. This occurred because a design change in the vessel made by the NSSS contractor failed to be coordinated with the site engineer. A major rework was required to place the cavity mounts in the ) roper )
position. This condition was determined to be unreportable ay the i
licensee and accepted as such by the NRC; therefore, I did not attempt to determine if other rework was required. '
My point in documenting this as a violation was to bring attention to the fact that the licensee had apparently failed to ensure proper installation of an important safety system plus cause a look to be taken at other safety systems to determine if the same problem exists.
j
- 3. Safety Significance .
As previously mentioned, problems had been encountered with the reactor '
vessel installation.n for both units. On these particular examples, it was !
almost impossible not to find the problem with the mounts and a fix was mandatory in order to set the vessel in place. The other problem was.not as apparent and could have remained undetected. One concern is what other q problems may have been created while correcting the identified situations. !
A more general concern is what problems may remain undetected because of a lack of QA presence. 1
- 4. Understanding of How Issue Was Randled by Management This item was documented as a violation in the draft report and changed to an unresolved item by management. This change was not discussed with me I in detail. It may have been discussed in more detail with Phillips. My l understanding was that management was of the opinion that the applicant did not have to audit the vessel setting because Criterion XVIII did not state specifically that the vessel must be audited. They felt that there i was no basis for a finding unless the licensee audit plan showed a scheduled audit and did not perform it. I'm not aware whether or not ,
consideration was given to past problems and the fact that no activity involving the vessel installation of either unit had been audited.
l ENCLOSURE.4 Inspection Report 85-07/05 !
Issue No. 4 (Attachment I to Attachment M-M)
- 1. Definition of Issue
" Spool Piece 301 (DWG No. BRP-CS-2-RB-76) had neither been marked with the material specification and grade nor heat number nor heat code of the material."
- 2. f Understanding of Issue My understanding of this issue was based on information gathered g the time of the inspection and subsequent discussions with the licensee and his contractors prior to the time of the OIA investigation. Inspection criteria and guidelines were based on the following:
- a. Regulatory Requirement - 10 CFR 50 Appendix B, Criterion VIII
" Measures shall be established for the identification and control of materials, parts, and components, including partially fabricated assemblies. These measures shall assure that identification of the item is maintained by heat number, part number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. These identification and control measures shall be designed to prevent the use of incorrect or defective material, parts, and components."
- b. ASME Code Requirement (Imposed by 10 CFR 50.55a and committed to in the F5AR, Table b.2-1) - Section III, 1974 edition, Article'NA-3766.6 requires that the identification of material consist of marking the material with the applicable material specification and grade of material, heat number or heat code of the material, and any additional marking required by this section to facilitate traceability of the reports of the results of all tests and 4 examinations performed on the material.
- c. Licensee Commitment -
(1) TUGCo QA Plan, Section 8.0, Revision 0, dated July 1, 1978, implements 10 CFR 50, Appendix B, Criterion VIII.
(2) Brown and Root Instruction, QI-QAP-11.1-26, Revision 4 dated August 28, 1980, requires that markings on bulk materials, such as pipe, be transferred prior to the material being cut.
Verification is required to ensure that markings comply with' Appendix 6.9E, Section 3.15 which in turn states that Q-pipe be marked with heat number, piece number, spool number, ASME grade number and schedule.
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- d. ' Inspection Guidance - no specific guidance is included in IE inspection modules.
This 3-inch dia_ spool piece was selected to be inspected.in order to verify that the site construction contractor was complying with
-appropriate requirements'for site fabricated items. The item was fabricated on site from bulk material and installed by the site contractor. NA-3766 is appropriate'in this instance and was implemented by the above instruction. A site code' engineer verified that they were working to this particular code. It is. recognized that Criterion VIII only requires that some means be provided to {
ensure traceability of items without being specific. The code; however, is specific and states certain marking requirements.
NA-3766 states that materit.1 requiring a CMTR (nominal pipe size greater than 3/4 inch will be marked with applicable specification and grade of material and heat number or heat code and that when
. material is divided, the identification marking'is required to be
' transferred to all pieces. Initially when this item was inspected, no markings couTd' be found by either myself or the site engineer who had accompanied me to the spool piece. Sometim later, I returned with another site engineer who pointed'out the spool piece number and drawing' number. Since the code requirements, which had been imposed by 10 CFR 50 Appendix B, had not been complied with in respect to the material ~ specification and grade and heat number or code I considered this violation and documented it as such-in the draft inspection report.
- 3. Safety Significance The significance of this finding in two-fold:
- a. A sample of one spool fabricated onsite is a very small sample and when problems are found, one can assume that the inspector stumbled onto an isolated case and that no other problems exist or on the other hand that a generic problem may exist, i 1
- b. If material was not adequately mat'ked prior to the cutting operations j the possibility exists that traceability has, been compromised and the f material is not actually what it is believed to be and the wrong material could be installed. la The above possibilities will never be explored if the finding is not' documented.
- 4. Understanding of How Issue was Handled by Management This issue was identified during the April 1 through June 22, 1985, period and documented as a violation in the inspection report. Several months
[ after the inspection had been completed, I was approached by Westerman who stated that he did not believe the finding was valid and requested that the issue be reviewed again. I reviewed the code and looked at the spool
3 piece, again, but still did not find all of the required markings. I told Westerman to delete the item from the report if that's what he wanted, since my opinion of the code requirements did not seem to matter. I also felt that by reviewing several records, I could establish traceability to the CMTR by using the spool piece designator, even though the intent of the code had not been met. My concern remained that the markings on the bulk material had not been transferred to sections being removed prior to ;
the cutting operation and that traceability h'da been-lost prior to the application of'the spool piece designator. .I discovered from OIA testimony by Barnes that Westerman had been provided information from Barnes pertaining to requirements of a vendor supplied spool piece rather.
than a site fabricated one. It was my opinion after reading the testimony, that once Barnes stated a position it became fact and no amount .
of discussion by me would have changed Westerman's decision. {
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ENCLOSURE 5 Inspection Report 85-07/05 i Issue 5 (Attachment 1 to Attachment M-M)
- 1. Definition of Issue '
" Failure to Perform Hydro test on Loop 3 Cold Leg prior to installation."
- 2. { { Understanding of Issue My understanding of this issue was based on information gathered at the~
time of the inspection and subsequent discussions with the licensH and hts cE tractors prior to the time of the OIA investigation. Inspection criteria and guidelines were based on the following:
- a. Regulatory Requirements - FSAR, Table 5.2-1, Commitment to ASME Code Section III, 1974 edition through Summer 1974 addenda.
- b. ASME Code Requirement (1) NB-6114(b) states in respect to the time of testing piping subassemblies, that the component test, when conducted in accordance with the requirements of NB-6221(a) shall be acceptable as a test for piping subassemblies.
l (2) NB-6221(a) states that the completed components shall be L subjected to a hydrostatic test prior _t_oo installation in the 1 system.
- c. Licensee Commitment - Brown and Root Procedure CP-QAP-12.2, Revision 8, dated June 11, 1984, paragraph 2.4 which addresses pressure testing of vendor items during system pressure testing states, in part: "When allowed by the Code, vendor components which have not been pressure tested by the vendor shall be tested as part of the system pressure test."
- d. NRC Inspection Guidance - No specific guidance in IE Modules.
During the records inspection, it was found that the NPP-1 Form stated that no hydro test had been performed for the RCS cold leg.
In discussions with Westinghouse and Brown and Root personnel, the statement was made that it is industry practice to defer the partial hydro test until the system is hydro tested. QAP-12.2, which covers inspection procedures for hydro testing discusses vendor components which have not been pressure tested, but the discussion makes no mention of piping subassemblies. It is, therefore, not certain that the discussion includes inspection of piping assemblies.
Based on the ASME Code referenced above, it is my opinion that deferral of the hydro for piping assemblies is not permitted.
NB-6114(b), which establishes the time of piping subassembly hydro
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testing refers specifically to NB-6221(a) which states " prior to installation." If NB-6221, in whole, was intended language such as that in NB-6114(c), which states that hydro testing of components and l appurtenances are to be performed in accordance with NB-6221, would '
be used. Additionally, both the NPP-1 Form was signed and it was ,
stated by contractor personnel that the code stamp had been applied '
to the subassembly prior to the hydro test. ASME code subsection NA-8230 states that the code symbol (stamp) is to be applied only after the hydro test has been satisfactorily completed.
This could lead field personnel conducting systems hydro tests to believe that the piping subassembly had been shop tested ,
The deferral of piping subassembly hydro testing seems to be based on opinions of what has been industry practice instead of requirements stated by the code which are requirements of 10 CFR 50.55(a). I do not believe that I, as an inspector, nor Region IV management has the authority to deviate from those requirements without specific written l instructions from NRC headquarters. Enforcement should be uniform throughout NRC and for this reason I documented the hydro test issue as unresolved rather than a violation in the draft inspection report pending an official resolution by NRC headquarters.
- 3. Safety Significance l The NRC inspection consisted of reviewing a very small percentage of class 1 piping. Serious questions arose as to the validity of tests being :
performed on the piping. First, code requirements were not followed, but the code data report had been signed and the code stamp applied, indicating that the subassembly had met all code requirements. It was ,
stated on the data report that no hydro test had been performed; however, this form was probably reviewed only by receiving inspection and the presence of the code stamp on the subassembly would carry more weight in the field where personnel could assume that the subassembly had been adequately sho) tested and inspect only items or welds which had been installed in t1e field. Also of concern is the generic implication of using an apparent industry practice rather than regulatory requirements in construction of the plant. How widespread is this practice? Probably, of most importance is the adequacy of the hydro test performed as part of the system test. In reviewing these records, it could not be determined that an adequate test had been performed. In questionawas whether or not material base metal repairs and shop welds had been adequately . inspected.
In order to get a better understanding contacted Mr. Kelley (Operations SRI)g , whoof thewitnessed had scope ofpart theofsystem the hydro, I tests, for his opinion of what was being inspected. He stated that he thought only field welds were being looked at. This issue was documented in the draft inspection report and the final inspection report as unresolved pending further evaluation by the applicant.
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{ Understanding of How Issue was Handled by Management
'I was approached by Westerman several months after this issue surfaced l suggesting that the issue be deleted from the report since Barnes, based j on vast experience in the piping industry, stated the deferral of piping l assembly hydro . testing was industry practice. We had several discussions, a including one meeting with Eric Johnson and other members of management {
where the issue was discussed. I'could not agree that an industry 1 practice, which seemed to conflict with code requirements, was acceptable 'I without documentation from NRC headquarters. 'I proposed at the meeting i that Barnes, an independent third examinethemeaningofN8-6114(b) party,andmyselfdiscussthecodeand andHB-6221(a). Barnes seemed very insulted at the. suggestion and we never had the discussion. At the meeting, management collectively agreed to delete the finding from the report and permit the inspector to pursue the matter through channels with headquarters. Westerman and I later discussed the matter, Bosnak at headquarters, who supported Region IV management. I requested this opinion in writing from Bosnak. I was prepared to make the request in.
writing but was informed by Johnson that a written clarification was not necessary and that perhaps I should receive additional code training. .I took no further action in the matter.
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i ENCLOSURE 6 Inspection Report 85-07/05 Issue No. 6
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- 1. Definition of Issue l "Sandusky Foundry and Machine Company test report for the cold pipe {
certifies that material meets requirements of ASME Section II, 1974 l editions through Winter 1975. Southwest Fabrication and Welding Company J I
code data report NPP-1 Form certified that the cold leg subassembly met requirements of ASME Section III, 1974 edition through Winter 1975. The l FSAR commitment is ASME Section III,1974 edition through Summer 1974." l Understanding of Issue
- 2. (
My understanding of this issue was based on information gathered at the time of the inspection and subsequent discussions with the license and his cE tractors prior to the time of.the OIA investigation. Inspection {
criteria and guidelines were based on the following: 1
- a. Regulatory Requirement - 10 CFR 50.55(a), issues dated both
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February 16, 1974, and January 1, 1985, state that the reactor j i
coolant presence boundary piping is required to meet the code edition and addenda in effect on the date of order of the piping.
- b. FSAR Table 5.2-1, dated March 30, 1979, lists the applicable code and q aodenda for reactor coolant pipe as ASME III, 1974 edition, through ;
Summer 1974. ]
- c. FSAR Table 3.2-1, dated October l'b, 1985, lists quality standards for Class 1 piping per NB-3600. A footnote to that table states that piping which are a part of the reactor coolant pressure boundary meet the requirements of the 1971 code with addenda through and including ^
Winter 1972. The note continues to state that later code revisions may be used optionally in accordance with requirements of 10 CFR 50.55a.
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- d. Licensee Commitment - Brown and Root QA Manual dated September 17, 1981, and revised June 11, 1984, states that the CPSES project shall be constructed in accordance with ASME Section III, Division 1, 1974 edition, including the Summer 1974 addenda for piping. Approved exceptions to use later codes were listed, but'were not applicable to Class 1 piping.
- e. NRC Inspection Guidelines - IE Module 49051-03 states that applicable portions of the SAR should be reviewed to determine licensee commitments and that the inspector should then use these SAR sections during the review of the licensee's implementing construction specifications, drawings, work procedures, and QA i. implementing procedures.
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- l. While performing an inspection of Unit 2 RCS boundary, during l A)ril 1985, it was observed that different sections of the FSAR and i tae construction contractor QA manual appeared to conflict. It was also observed that the Code Data Report (NPP-1) and CMTR for the cold leg piping was to ASME Section II/III 1974 edition through ,
Winter 1976 addenda, which conflicted with both FSAR and QA manual i commitments. Since further review of this situation was required to )
determine the appropriate code which should have been used and =
whether or not a violation existed the issue was documented in the j inspection report draft as unresolved with action required by the i licensee for further evaluation. ]
- 3. Safety Significance Without further evaluation, the safety significance of this issue cannot be fully determined. It was cumented as unresolved for that reason.
4 L Understanding of H2w ftgue was Handled by Management j Several months after this issue had been drafted as an unresolved issue I i was approached by Westerman about removing th'e item from the report. The draft had apparently been reviewed by NRC headquarters who indicated that the issue should either be a vfolation er stricken from the report.
Westerman and Barnes were of the opinion that any code issue which had been approved by 10 CFR 50 M was acceptable. I could not agree with their opinion and, therefore, could not delete the item. I stated to' Westerman that if he wanted the item deleted that he would have to delete it himself. The item was deleted and was to be pursued with NRC headquarters separate from the report. I do not know the status of that effort; however, Amendment 57 to the FSAR, dated December 20, 1985, changed the code requirements for RCS piping to ASME III, 1974 edition through Winter 1975 which agrees with the material CMTR, however, there is no documentation of the above discussions in the NRC inspection reports.
Additional code concerns similar to the above, including Class II piping l which did not comply with licensee commitments were not pursued because of ,
NRC management objectives, which in one instance was displayed by Barnes during an exit meeting with the licensee. .
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ENCLOSURE 7 Inspection Report 85-14/11 Issue No. 7 and 8 (Attachment 1 to Attachment M-M)
- 1. Definition of Issue
" Site records of Chicago Bridge and Iron shipped to Houston, Texas, in cardboard boxes and no backup copy of records were made."
Understanding of Issue
- 2. (
My understanding of this issue was based on information oathered at the time of the inspection and subsequent discussions with tIIs licensH and his cli6trE tors prior to the time of the OIA investigation. Inspection criteria and guidelines were based on the following:
- a. Regulatory Requirement - 10 CFR 50, Appendix B, Crtterion Ii (1) "The applicant shall establish . .*. a quality assurance program which complies with the requirements of this appendix. This program shall be documented by written policies, procedures, or l instructions and shall be carried out throughout plant life in 4 accordance with those policies, procedures, or instructions."
(2) FSAR, Paracraph 17.1.17 states, in respect to QA records that TUGCo and its contractors have a quality records system which consists of procedures to protect records against deterioration and damage. Appendix 1A(B) of the FSAR commits to Regulatory Guide 1.88, ANSI 45.2.9-1974 and ANSI 45.2.9-1979 which permit the following two methods of protecting QA records from possible destruction: (a) permanent or temporary facilities capable of providing protection to the records and (b) maintaining duplicate records stored in a separate location.
- b. Licensee Commitment - Chicago Bridge and Iron Nuclear Records Procedure (NRP-1), Section 5 provides instructions for packing and It shipping records to the Nuclear Records Center at Houston, Texas. j is stated that corrugated cardboard packing is satisfactory for small shipments and that wood should be used for large freight shipments.
Shipping by Air Freight, Air Express or UPS is required for shipments containing documents difficult to replace.
During preparation fer an inspection of QA reco,-ds for Unit 2 mechanical penetrations it was discovered that the original site generated construction records for the penetrations and containment liner had been sent to CB&I Houston, Texas, office for reproduction.
The matter of how records were shipped was discussed with the site supervisor who was in charge of the CB&I contract. I concluded that records were shipped in accordance with the C6&I procedure and did not pursue the matter at this time due to other priorities. Several
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2 .l weeks passed and another inspector found that' records generated by another site contractor had been shipped offsite also. As a result ,
the SRI' asked me to look deeper into the CB&I matter.- l
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I had further discussions with the subcontract administrator and-reviewed the CB&I records. procedure (NRP-1). To the best of my knowledge the records were shipped in cardboard ~and wooden boxes as L
permitted by the procedure. I had been tasked with getting the CPSES tracking system in place, which took priority over inspecting, therefore, .I submitted n1y finding with a- recommended violation to the SRI. The recommended violation cited the licensee for failure'to establish procedures to control QA records in accordance with FSAR connitments.
During the time of the inspection, I discussed the . issue with the licensee assistant QA Manager who informed me that the records were-
- shipped at' their (licensee)' risk and -if anything happened to the j records that they would have to be reconstructed.
My understanding in this issue.in that records generated onsite, whether by.the construction contractor or a subcontractor, are site records and are to be controlled by procedures which comply with the FSAR. Just because' ANSI'45.2.9 does not specifically address records' in shipment does not relieve the' licensee from complying with one of the two ANSI provisions for record protection. Even'if the records are assumed to be' vendor records adequate protection would normally ,
be provided since the vendor maintains a copy of each record shipped. I I alsoicould not accept the argument of reconstructing records. In addition'to parts of the liner being buried in concrete, it would be impossible to retrieve information such as CMTR's,-weld material records, welding operator records, and information on construction travelers. Even when records arrived back onsite it was impossible to determine accountability since no inventory was taken of the 1 records prior to shipment.
During April 1986, after records had been returned to the site discussion was held, concerning the issue of record accountability, !
with a licensee QA supervisor. He stated that upon return some boxes were labeled with contents and others were labeled as " records" and that the licensee had reviewed purchase orders to determine what types of records were required. It was determined that each type of record had been returned, but not that' all records of each type had been returned. Without an inventory it; therefore, could not be determined that all records were returned.
- 3. Safety Significance - I cannot comment on this aspect of the issue without addition evaluation.
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' 4. Understanding of How Issue was H&ndled by Management At the time this issue was documented in an inspection report, I was involved in establishing a CPSES tracking system and was not performing inspections.- My recomended input was coordinated with the SRI and I'm unaware of how management handled the issue.
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