IR 05000458/1981009
| ML20031G729 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 10/05/1981 |
| From: | Beach A, Crossman W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20031G728 | List: |
| References | |
| 50-458-81-09, 50-458-81-9, NUDOCS 8110230551 | |
| Download: ML20031G729 (18) | |
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APPENDIX
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U. S. NUCLEAR REGULATORY COMMISSION 0FFICE OF IN5PECTION AND ENFORCEMENT
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REGION IV
Report:
50-458/81-09 Docket:
50-458 Licensee:
Gulf States Utilities Post Office Box 2951 Beaumont, Texas 77704 Facility Name:
River Berd, Unit No. 1 Inspection at:
River Bend Site Inspection Conducted:
July 16 through August 31, 1981
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Inspector:
V (00 l
enrA; 4/lo/Al A. B. freach, Resident Reactor Inspector IDate Projects Section 3 Approved:
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W. A. Crossman, Chief, Projects Section Date Inspection Summary:
Inspection During Jsly ant August (Report No. 50-458/81-09)
Areas Inspected: Routine, announced inspection by the Resident Inspector (RRI)
including follow-up of previous inspection findings; follow-up of licensee identified items; activities with regard to site development work; concrete placement activities; welding of safety-relattd piping; safety-related piping support and restraint installation; mechanical equipment installation and in-place storage of equipment; structural steel erection; and Class IE electric equipment qualification. The inspection involved 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> by one NRC inspector.
Results: Of the six major areas inspected, no violations or deviations were identified in five areas, one violation was identified involving in-place storage (Violaticn-Failure to Follow Storage Procedures for Motor-Operated Valves).
8110230551 811009 PDR ADOCK 05000458
G PDR
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DETAILS 1.
Persons Contacted Principal Licensee Employees
- P. D. Graham, Director, Quality Assurance
- C. L. Ballard, Supervisor, Quality Assurance R. B. Stafford, Supervisor, Quality Assurance-G. V. King, Supervisor, Quality Assurance
- H. G. Domschke, QA Engineer
- W. S. Stuart, QA Engineer
- E. A. Troncelleti, QA Engineer
- T. C. Crouse, Superintendent, Site Construction
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A. Walton, Director (Acting), Site Engineering Stone and Webster Personnel C. D. Lundin, Manager, Project Quality Assurance
- R. L. Spence, Superintendent, Field Quality Control (FQC)
- G. d. Byrnes, Assistant Superintendent, FQC
-*1 D. Davis, Assistant Superintendent, FQC
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- R. L. Whitley, Assistant Superintendent, FQC
- W. I. Clifford, Resident Manager
- E. A. Sweeny, Superintendent of Site Ensineering
- P. D. Hanks, General Superintendent, Construction
- D. P. Barry, Superintendent, Construction Services The RRI also interviewed additional licensee, Stone and Webster, and other contractor personnel during this inspection-period.
- Denotes those persons with whom the RRI held on-site management meetings during this inspection period.
2.
Licensee Action on Previous Inspection Findings (Closed) Violation (50-458/81-06):
Inadequate Acceptance Criteria For In-Place Density Testing of Backfill. The River fend PSAR (Preliminary Safety Analysis Report) defines that the soil stratiography at the
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. site for engineering purposes, may be subdivided into five well defined zones:
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Post Hickey Top-Stratum Silts and Clays
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Sands and Clayey Sands
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Citronelle Buried Channel Deposits, Sands, and Gravels
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t Tertiary (Pascagoula) Clays
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l The buried channel sand and gravel depc;its of the Citronelle are the primary natural founding strata for the seismic Category I structures,
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An analysis of the PSAR indicates that these deposits below the proposed excavation levels have a relative density generally greater than 70
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percent with only a few values less than 60 percent randomly scattered throughout the deposits.
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In Section 2.5.4.5, " Excavation and Backfill," the PSAR states, "It is planned that all seismic Category I structural backfill be compacted to obtain 70 percent relative density (with minimum acceptable relative density being 60 percent)."
The River Bend FSAR (Final Safety Analysis Report), while tendered by the applicant but not docketed by the NRC at the time of that inspection period, states, "The highest value of dry density corresponding to 70
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percent relative density was chosen as the minimum acceptable in place density." But, as stated in the violation, from the review of the
in place density test reports, the NRC inspector determined that the unit weight value listed as the minimum acceptable density corresponds to 60 percent relative der.sity, representing blanket approval of the 60 percent relative density value as being the minimum acceptable.
Minimum relative density also has an effect on the concept of liquefac-tion.
In the " Foundation Engineering Handbook," edited by Winterkorn and Fang, D'Appolonia suggests that liquefaction might occur in
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cohesionless soils of relative density less than 50 percent during ground accelerations of approximately 0.lg: while for relative densities
greater than 75 percent, liquefaction is unlikely for most earthquake loadings.
(The maximum vibratory acceleration associated with an-earthquake at the River Bend site would be 0.07 ; however, it is
assumed to be 0.lg, since this is the minimum acceptable value in accordance with 10 CFR 100, Appendix A.)
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The response by the licensee, however, indicates the acceptance criterion for in place density was chosen with conserwatism such that the resulting level of compaction is generally in excess of 70 percent relative density as determined by complete testing.
In fact, the 1;censee further states the average measured relative density is 94 percent, with only
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five test results between 65 and 70 percent.
No relative density test was below the 65 percent value for any of the Category I backfill material.
The NRC inspector reviewed ten relative density tests performed during April and May 1981. Values for the relative density tests ranged from 82 percent to 106 percent.
The average relative density value for these ten tests was 93 percent.
Thus, since Category I backfill is essentially
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complete, the licensee's response is, in fact, in accordance with their PSAR commitments; that is, seismic Category I backfill be compacted to a relative density of 70 percent, with the minimum acceptable density being 60 percent.
With respect to liquefaction, the results of the analysis presented in Tables 2.5.4-7, 2.5.4-8, 2.5.4-9, and 2.5.4-10, indicate a minimum
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numerical factor of cafety of three with respect to initial liquefaction.
Since actual test values for relative density are higher than the test values used in the analysis, the backfill has a larger margin of safety against liquefaction than demonstrated.
Thus2 since Category I backfill is essentially complete at the time of this inspection, this item is considered closed.
(Closed) Unresolved Item (50-458/81-07):
Concrete Density Requirements Within the Primary Shield Wall.
The NRC inspector met with licensee and contractor rep;esentatives on July 23, 1981, to discuss his concerns identified relative to the proper qualification for use of the heavy-weight aggregate concrete mix used in the Primary Shield Wall.
The first concern di; cussed in this meeting involved the acceptance criteria applicable to the lower one-third of the Primary Shield Wall.
This minimum density acceptance criteria was 135 pcf, for biological shielding.
The Stone and Webster representatives stated that the shielding engineers determined that " normal weight" concrete may be used in this region as it was not a part of the active core region, and thus, not a requisite for biological shielding.
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However, the actual density of the (wet) mix utilized in this area was 209 pcf.
The accept / reject criteria for the first third of the Primary Shield Wall used during the placement of the mix, was stated i
to be 200 pcf (dry), in accordance with the River Bend PSAR requirements.
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A Stone and Webster Engineer & Design Coordination Report (E&DCR) C-2719 was issued on May 7, 1981, indicating a revision of 5 tone and Webster Specification P10.350, Revision 3, " Mixing and Delivering Concrete,"
incorporating *..a 200 pcf minimum dry der.sity requirement and a 208 pcf minimum wet density requirement.
The report states that "the radiation protection analysis shows that an air dry unit weight of 200 pcf provides adequate shielding for the Primary Shield Wall."
However, this revision was made after placement of concrete in the lower one-third of the Primary Shield Wall.
The second concern addressed qualification of the concrete mix in accord-ance with ACI 211, "Recommendeu Practice for Selecting Proportions for Normal and Heavyweight Concrete." An " Examination and Mix Design for the Higt Density Fill Material for the Primary Shield Wall," dated March 27, 1981, shows the results of the preliminary and prequalificattun tests performed on the material.
An analysis of the grout mix design components was performed by the Pittsburgh Testing Laboratory in accordance with ASIM C-114.
This is documented in Report 814025, dated March 24, 1981.
In addition, a review of Report 812796, dated December 18, 1980, " Report of Hematite Fine Aggregete," indicated that the fine aggregate material used in the concrete mix was in accordance witt the requirements of Section 4.1 of ASTM C-637, " Standard Specification for Aggregate for Radiation-Shielding Concrete."
The third concern addressed by the NRC inspector in NRC Inspection Report 50-458/81-07 involved the shielding calculation for the lower one-third of the Primary Shield Wall.
However, since it could be established from documentation reviewed that the actual in place density of the r.oncrete exceeded the minimum PSAR requirements, a review of this calculation was not ree" ired.
The heavyweight concrete aggregate should not have been placed in any location prior to establishing firm accept / reject criteria.
However, this problem was properly identified in accordance with the licensee's Quality Assurance Program, properly documented, and properly addressed by the responsible individuals.
This item is considered closed.
(0 pen) Unresolved Item (50-458/81-06):
Substitution of Grade 60 Rein-forcing Steel for Grades 40 and 50 deinforcing Steel.
This information has stili not been obtained.
If this information is not available within the next inspection period, this item wi'l be upgraded in accordance with the Interim Enforcement Policy, 45 FR 66754 (October 7, 1980).
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Licensee Identified Construction Deficiency Reports (0 pen) Miscellaneous Steel Provided by LIVES Steel.
The information requested in NRC Inspection Report 50-458/81-06 has still not been re-ceived; i.e., that assurance that the design calculations for the mat anchor ring reflect that the actual stresse are less than the mar.imum allowed and that the appropriate design criteria, as defined in the safety analysis report, has not been violated by the low yield strength readings obtained during tensile testing.
Additionally, assurance is required that none of the suspect material performs an actual structurai function
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or has little or no structural significance.
This item remains open.
l The NRC inspector reviewed River Bend Project Procedure RBPP 1.7,
" Processing 10 CFR 50.55(e)/10 CFR 21 Reportable Deficiencies, Defects,
and Noncompliances." The stated purpose of this procedure is to provide
... a method for initiating, reviewing, and dispositioning reportable
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deficiencies, defects, and noncompliances." Stone and Webster Quality i
i Standard QA-16.2, Revision 0, " Notifying Clients of Potentially Reportable Deficiencies Under 10 CFR 50.55(e)" was likewise reviewed.
10 CFR 50.55(e)(2) states, "The holder of a construction permit shall within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notify the appropriate Nuclear Regulatory Commission Inspection and Enforcement Regional Office of each reportable deficiency."
This requirement was discussed with licensee management as dccumented i
in IE Inspection Report 50-458/79-07.
Additionally, in a response to i
this infraction in IE Inspection Report 50-458/79-05 (Failure to Provide Timely Notification of Construction Deficiencies), the licensee states, "In the future, Gulf States Utilities will inform the MRC Regional Office by telephone call within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when a problem or
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deficiency is discovered that could possibly be a reportable deficiency, e
but adequate information will not be available to determine the report ~
ability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery of the problem or deficiency."
However, once a deficiency is identified, the Stone and Webster pro-cedure allows 21 days to complete the cycle for a "Potentially Report-able Deficiency." The River Bend Project Procedure does not establish
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time requirements for review to reportability.
Discussions with licensee representatives indicated, to the NRC inspector, that there still may be some individuals who are misinterpreting the
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requirements.
The Senior Vice President for the River Bend Nuclear Group has repeatedly emphasized that if there be any doubt, report the deficiency promptly.
However, to ensure that all licensee representatives
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responsible for construction deficiency reporting are aware of their appropriate commitments, the following was suggested to the licensee, as taken from the text of IE Inspection Report 50-458/79-07.
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Telephone the regional office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the
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deficiency is considered reportable based on facts available within the 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> period.
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Indicate thero may be a need for more time for testing and i
analysis, c.
If preliminary results indicate the material is not acceptable,
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then extensive evaluation and/or rework may be required.
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so, it is clearly reportable.
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If the item had not been promptly reported to NRC, an NRR I
or IE evaluation of the possible impact on the design and construction of other facilities could not be made in a
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timely manner.
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e.
If tests and analysis indicate there is no problem, then a
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subsequent 30-day written report is not required, and a simp 16 letter to RIV stating this fact is all that is
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required.
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The licensee must maintain records to demonstrate that an adequate evaluation / analysis was made regarding the impact on safe operations.
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i Written reports are required within 30 days.
Context of the report is j
stated in 10 CFR 50.55(e).
Interim reports must state the date when the final report or the next interim report will be issued.
The licensee was urged to discuss reportability with the responsible IE
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inspector if there is doubt or if questions exist.
The ultimate responsi-
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bility for reporting is with the licensee.
If the licensee were to j
habitually evaluate events as non-reportable, the NRC would have to take
appropriate corrective action.
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Using this guidance, the licensee can provide prompt notification based
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l on facts available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and still maintain the option to
clarify the significance of the matter as further information is developed.
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The NRC inspector re-emphasized that the ultimate responsibility for re-porting is with the licensee.
Both on-site and off-site engineering reviews and nonconformance reports were reviewed in this inspection period and will continue to be reviewed in the next inspection period to ensure the requirements of 10 CFR 50.55(e) are being satisfied.
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4.
Site Tour The NRC inspector toured the safety-related plant areas several times during the inspection period to observe construction progress and the general practices involved.
Fire prevention and fire protection practices were observed.
During one of the tours, in place storage of the Reactor Vessel was reviewed.
The Stone and Webster Specification 229.170, Revision 3,
" Storage and Maintenance of Permanent Plant Equipment," establishes the present requirements for storage of the reactor vessel.
A discussion was also held with Field Quality Control personnel.
Their inspection plan to Stone and Webster Specification 229.160, " Mechanical Installa-tion-Category I,"
also encon. passes requirements for n.onitoring of the storage of the Reactor Vessel.
The warehouse storage facilities at the River Bend site were also toured.
During this tour, the NRC inspector noted several problems involving the storage requirements.
The first concern identified seven notor-operated valves whose space heaters were not energized.
(These are different from those valves referenced in paragraph 9.) A violation was issued on August 7, 1981, for those referenced valves in that their heaters were likewise not energized.
Contractor personnel stated that most of these had been electrically disconnected to facilitate the removal of other equipment out of the immediate area, and had not yet been reconnected.
The NRC inspector re-emphasized the procedural requirements to the cont" actor representatives, and will closely review the licensee's corrective action taken in response to the identified violation (paragraph 9).
Another concern identified was the storage of items to allow for equip-ment access for inspection or maintenance.
Timeliness for the initiation of Equipment Storage History Cards (as required by Specification 2291^/0, Revision 3) and the performance of visual inspections were also identified.
A Stone and Webster Type C Inspection Report, X1000410, dated July 17, 1981, identified similar nonconforming conditions (with the exception of energized space hearters for motor-operated valves).
Contractor Field Quality Control emphasized corrective action was being implemented to correct these deficiencies identified in the above inspection report.
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Space availability is becoming a problem in that the available warehouse
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space at the River Bend site is somewhat limited in relation to the
amount of equipment being reces.ed.
This problem was discussed in IE
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Inspection Report 50-458/81-02.
The licensee is attempting to locate additional off-site space to accommodate this equipment.
The NRC inspector emphasized to licensee and contractor personnel that problems involving storage space should not and must not jeopardize the
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i required maintenance or level of care that must be maintained to i
ensure the quality of the equipment necessary to perform its safety function. Although not viewed as a problem at the time of this inspec-tion period, the allocation of space must be closely monitored to ensure
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i that all safety-related equipment is properly stored and maintained, and
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that maintenance and other activ?
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t The NRC inspector also placed special emphasis on the review of Field
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Document Control.
A meeting with licensee and contractor personnel
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was held on kJgust 13, 1981, discussing the Document Control Program.
Contractor personnel indicated their confidence in the Program and g
were satisfied that their program is now being implemented effectively.
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Findings relative to the Document Control Program were not identified
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during this inspection period.
The Contractor's Document Control
Program, as well as the licensee's corrective action taken in response to their findings involving the Document Control Program, will contirue j
to be reviewed to ensure that the applicable requirements have, in fact, been satisfied.
Thus, the unresolved items identified in IE Inspection Report 50-458/80-13 (" Construction Verification Program") and in IE Inspection Report 50-458/81-02 (" Document Control") will remain open.
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No violations or deviations were identified.
5.
Site Development Work
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The NRC inspector reviewed the applicable portions of the PSAR in con-junction with the requirements set forth in 10 CFR 100, Appendix A, Section VI.
Specification 210.100, Revision 2, " Site Development
i Work," establishes the requirements for Category I backfill operations at the River Bend site.
A sieve analysis, in accordance with ASTM C-136-71 (except for Section 5.3 as modified in the specification), is required to l
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l be performed by this specification, as well as a maximum / minimum density test reouirement in accordance with ASTM D2049-69 using an electromagnetic vibratory table.
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Fifteen inspection reports were reviewed during the period of April and May 1981.
The 15 reported sieve analyses and the 10 reported maximum /
minimum density cests were found to be in accordance with the soils testing requirements of Specification 210.100, Revision 2.
Testing frequencies met the specified requirements.
No violations or deviations were identified.
6.
Concrete Placement Activities The NRC inspector reviewed placement records, batch tickets, and curing records for the following concrete placements during this inspection period:
Date
Inspection Report
Placement 5/01/81 IRS 1002213 RB-1-P82-5 5/26/81 IRS 1002638 PWS-HDFM-125 6/11/81 IRS 1002891 RB-DW-TD-95 7/03/81 IRS 1003262 RB-3-150 7/11/81 IRS 1003377 AB-8-W109G2 7/14/81 IRS 1003424 RB-7-W99 7/24/81 IRS 1003611 PSW-HDFM-125 7/26/81 IRS 1003638 RB-7-W105 8/06/81 IRS 1003848 A6-10-W128G5 Portions of the placement activities performed on RB-7-W105 and AB-10-W128G5 were observed.
All records reviewed and activities observed in-dicated proper consistency of placement, timeliness of concrete mixer truck batching to discharge, and implementation of proper concreting practices. Qualification of inspection personnel were reviewed and found to be in accor dance with the established site requirements.
All records reviewed and activities observed were found to be in accordance with site Specification 210.370, Revision 6, " Placing Concrete and Reinforcing Steel."
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The site Batch Plant was toured with Stone and Webster Field Quality Control civil supervisory personnel. Aggregate storage and batch plant operations were observeo. All activities observed were found to be in accordance with Stone and Webster specification " Mixing and Delivering Concrete." The Stone and Webster personnel discussed improvements made in the batch plant operatiw to improve concrete consistency, and gave a short summary of their day-to-day activities performed at the Batch Plant.
No violations or deviations were identified.
7.
Welding of Safety-Related Piping Shop fabrication and field erection of piping is in progress. As was discussed in IE Inspection Report 50-458/80-13 (paragraph 5), the site weld-to weld r9 ject rate is again beginning to rise. Licensee and con-tractor personnel stated in discussions with the NRC inspector that a likely cause for this condition was the initiation of field welding.
However, licensee representatives also stated that they were evaluating courses of action necessary to ensure quality welds and as low a weld reject rate as is possible in accordance with their connlitment to meet Criterion 14 of Appendix A to 10 CFR 50 for Class 1 welds. Although welding activities have been in piogress for some ten months, less than two percent of the entire safety-related welds have been completed at the time of this inspection.
The NRC inspector will review the licensee actions taken relative to the welding program to ensure that proper action is being taken to minimize weld repairs.
No violations or deviations were identified.
8.
Sarety-Related Piping Support and Restraint Installation The NRC inspector initiated this portion of the inspection with a review of documentation relative to matericl used in the manufacture of safety-related component supports. This documentation included Nonconformance and Disposition Report (N&D) A-001, dated May 23, 1981.
This N&D indicated that ASTM A-500 Grade B tubing has been procured from a supplier who was not qualified in accordance with the vendor's Quality Assurance (QA) program, in violation of the ASME III QA program. Addition-ally, the vendor did not perform any confirmatory tests on the material prior to its acceptance.
The disposition to the N&D allowed acceptance of the ASTM A-500 Grade B tubing prc ured from the supplier and allowed shipment of the material to the River Bend site upon completion of a confirmatory testing program.
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Subsequently, on June 8,1981, N&D A008 was written documenting that two of the seven referenced heat numbers of material tested as part of ti.e
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confirmatory testing program did not meet the minimum specified elonga-
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tion requirements.
Retests from the same material aiso failed.
Five heat numbers of material were documented as meeting the specification requirements; the remaining material was dispositioned for use in Class 4 supports.
Contractor representatives indicated that these conditions had been evaluated for reportability in accordance with 2t, JFR 21, but felt the condition was "a one time only occurrence."
i Further review into this matter revealed that another supplier of ASTM A-500 Grade B material would no longer verify material procured under their existing Quality Assurance Program at the time of purchase after April 15, 1981.
After this date, all material must be procured to their
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specification STN-50B (aluminum filled steel).
The supplier " believes that potential long term metallurgical problems might result from the use of ASTM A-500 Grade B material manufactured to commercially acceptable practice." However, this supplier was required to comply with 10 CFR 21, as part of its existing Quality Assurance Program, and no evidence of reportability could be given to the NRC inspector.
Thus, the NRC inspector requested in a memorandum dated August 31, 1981, to the Vendor Inspection Branch in tl.a Region IV Office of Inspection and Enforcement, that they review these conditions for applicability to i
10 CFR 21 requirements.
In addition, the licensee is currently reviewing this matter for reportability under the requirements of 10 CFR 50.55(e).
l However, it is important to note that the licensee has not reported this matter to the Region IV office at the time of this inspection (reference paragraph 3).
Until all of this information can be obtained, the reportability requit 1ents as to the use of this steel is considered to be unresolved.
Contractor and licensee representatives also identified a difficulty in
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procuring ASTM A-500 Grade B structural tubing and are evaluating proper procurement requirements for the purchase of this material in accordance with Appendix B to 10 CFR 50.
Until the NRC inspector can adequately de-termine whether or rot these requirements are completely satisfied, this matter of purchase requirement is also considered to be unresolved.
No violations or deviations were identified.
9.
Mechanical Equipment Installation and In-Place Storage of Equipment The NRC inspector reviewed in place storage and protection of mechanical
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equipment.
During a site tour on July 27, 1981, he observed that no space heaters were energized for four motor operated valves installed in place
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in the "D" tunnel. These valves are as follows:
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1-SWP-MOV-96A 1-SWP-M0V-968
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Additionally, valve 1-SWP-MOV-96A was installed beneath an open area in the Auxiliary Building Tunnel.
The cover over the valve was not properly
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affixed to allow proper air circulation and drainage.
Consequently, the l
limit switch was submerged in water trapped in the cover.
A subsequent megger check of the motor failed.
These conditions were subsequently documented in Nonconformance and Disposition Report No. 1550 by Stone and Webster Field Quality Control (FQC).
Specification 229.170, Revision 3, "Sto. age and Maintenance of Permanent Plant Equipment," states in paragraph 1.5.2, " Motor-operated valves shall have their motor heaters and limit switch compartment heaters energized while in storage."
In addition, paragraph 1.6.2 states, "... the area shall be provided with unifor.n heating and temperature control or its
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equivalent to prevent condensation and corrosion." Likewise, Section 3 j
of thic specification (page 3-38) requires that space heaters be energized and tnat Level B storage requirements be improvised.
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Criterion V of Appendix B to 10 CFR 50 requires that activities affecting quality shall be prescribed by documented instructions or procedures of a type appropriate-to the circumstances and shall be accomplished in accord-ance with these instructions or procedures.
Furthermore, Criterion XIII of Appendix B requires that measures shall be established to control the storage and preservation of equipment in accordance with instructions to
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prevent damage or deterioration.
Thus, these conditions are considered to
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be a violation of Criterion V of Appendix B to 10 CFR 50; i.e., failure to i
follow storage procedures for motor-operated valves.
The NRC inspector emphasized to licensee and contractor representatives that proper planning must be utilized in providing for in place storage to prevent similar problems from occurring in the future.
A memorandum from T. L. Harpster, Chief, Reactor Projects and Resident Inspector Branch,
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dated August 14, 1981, states, "The level of care for an item should, in general, be the same regardless of location.
That is, the level of care which is necessary to protect and maintain, or restore the quality of all f
i elements of the item important to its safety function.
Planning is impor-tant in determining when an item should be placed in its final location',
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what conditions will be experienced at that location, and which measur e are necessary to protect and care for the item."
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In addition. F NRC inspector reviewed vendor documentation packages in regard to mo' ' operated valves and their acceptance on site. The follow-ing inspection reports were reviewed:
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Inspection Report
Date M0100448 12/14/80 M0100461 12/20/80 M0100462 12/20/80 M0100467 12/21/80 M0100469 12/21/80 M0100472 12/22/80 M1100053 01/24/81 M1100055 01/31/81 M1100062 02/03/81 M1100093 02/16/81 M1100098 02/17/81 M1100185 03/28/81 M1100205 04/12/81 M1100224 04/25/81 M1100261 05/23/81 M1100277 06/02/81 M1100291 06/12/81 M1100319 06/26/81 M1100323 06/26/81 M1100370 07/18/81 M1100381 07/26/81 M1100386 07/26/81
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M1100401 08/09/81 M0100452 12/15/80 M1100327 06/27/81 i
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M0100466 12/21/80 M0100468 12/21/80 M1100013 01/13/81 M1100027 05/31/81 M1100290 06/12/81
M1100318 06/26/81
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A review of this documentation revealed that these valves were received
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on-site without ?pproved seismic calculations.
The River Bend PSAR states in Section 3.7.5.2A, on page 3.7-37, "... the actual vendor is not j
allowed to ship a component unless its seismic adequacy is reviewed and accepted."
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E&DCR-P-10198A, dated November 12, 1980, however, indicated that a PSAR chanse had been initiated on November 13, 1980, to allow shipment of the valves prior to receipt, review, and approval of seismic calculations by Stone and Webster Engineering.
The E&DCR also directed site Stone and Webster Field Quality Control to verify that seismic calculations have been submitted to the engineers and stamped " approved" in accordance with the requirements of the specification.
This disposition was likewise approved i
by the Stone and Webster Quality Systems Division on November 13, 1980.
A review of '.ne subsequent PSAR change dated November 13, 1980, reads as
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follows, "
the actual vendor is not allowed to ship a component unless
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its seismic adequacy is reviewed and accepted; or the vendor provides a
' Certificate of Conformance-Seismic' based on experience in satisfying the design of similar components for the same or higher seismic loads and accepts
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full liability for all modifications to be made as a result of final analysis."
A review of the July 29, 1981, " Licensing Document Change Request Log" indicated this change to still be under Stone and Webster review.
i One hundred twenty motor-operated valves were received at the site in accord-ance with E&DCR-P-10198A.
However, FQC personnel were unaware of their respon-sibility to verify that seismic calculations had been submitted and approved by their engineers.
Licensee representatives were unaware of the PSAR change in that it apparently was still under Stone and Webster review.
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Criterion VII of Appendix B to 10 CFR 50 requires that purchased materials conform to procurement documents.
Criterion VIII requires identification and control of materials and equipment.
During this inspection period, a program has been established by licensee and contractor personnel which defines the method employed on the River Bend Site to track engineering 6pproved equipment that does not meet seismic and enviormental qualification requirements at the time of instal-lation, but must be available prior to system turnover.
However, the NRC inspector needs to review implementation of this program to ensure that applicable purchased equipment is properly controlled.
Until the implementation of this program can be reviewed, this matter is considered to be unresolved.
10.
Structural Steel Erection The NRC inspector reviewed Stone and Webster Specification 210.350, Revision 6, " Structural Steel." The applicable PSAR requirements, as well as the Field Quality Control (FQC) inspection requirements for structural steel erection, were reviewed.
FQC structural inspection personnel were checked against the current FQC training matrix to ensure their proper qualification in accordance with their assigned tasks.
Installation of supplemental steel in the Control Building was observed; specifically, HVAC support DGR 7625 at Control Building Elevation 70.
Work was being performed in accordance with Drawing BZ539E-11 by welder S-01.
Weld material was documented on Weld Material Requisition 203705.
All work observed conformed to the applicable specification requirements.
Torquing of bolts for structural steel connections for Beams 4-3B1-1, 4-882 through 4-888 was observed and found to be performed in accordance with Bristol Steel and Iron Works Drawing CE 3, Revision 5, and Stone and Webster Sketch SK-S-2785-1.
In addition, an erection inspection of steel erected for the Control Building Elevation 120 was reviewed, as documented on Inspection Reports IRC 1200447 and IRS 1200508.
Welding of structural steel to embedded plates was observed and found to be in accordance with Bristol Steel and Iron Works Drawing AE 2, Revision 6, and Stone and Webster Drawing ES-66A, Revision 9.
Also, the NRC inspector selected several beams for appropriate measurements to ensure their sizes were as required by the as-built drawings and the applicable design parameters required by Part 1 for W-shaped Beams of the AISC Steel Construction Manual.
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Nonconformance and Disposition Report 1498 was reviewed for Beams AR-75, AR-73, and AR-76.
This nonconformance documented sniping corners for ex-cessive copes for there referenced beams.
Inspection Report 51201315, in-volving forging bursts relative to A-325 bolts was likewise reviewed.
Some 48 Inspection Reports involving bolt torquing for structural steel connections were reviewed, dated July 27, 1980 to July 7, 1981.
All records revieved were found to be in accordance with the N me and Webster Specification 210.310, Revision 2, " Structural Steel."
No violations or deviations were identified.
11.
Electrical Equipment Qualifications The licensee has committed to conform to the requirements of IEEE 323-1974 for enviromental qualification of Class IE electrical equipment.
Section 7.9 of the SER (Safety Evaluation Report), dated December 1975, states "this program requires that the exact method of qualification of each type of Class IE electrical equipment to IEEE 323-1974 be individually identified.
The program specifies that this information will be provided to the staff for review and approval as equipment vendors are defined.
In addition, a program for ongoing qualification was described and will be performed for equipment with a qualified life less than full plant life.
The specific details for each type of equipment will be provided to the staff for review and approval when the equipment is identified and the details are established.
In addition, Amendment 13, December 1975, to the Preliminary Safety Analysis Report (PSAR), describes the implementation program for the IEEE-1974 equipment qualifications (as discussed in IE Inspection Report 50-458/80-08).
Part A of this section states " Gulf States is working with its NSSS vendor
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and Architect Engineer to develop a complete listing of Class IE equipment i
for River Bend Station by type application, location, manufacturer, name, model number, part number, or similar identifying number."
However, some NSSS equ pment, possibly all NSSS supplied equipment, was manufactured to IEEE 323-1971.
Licensee representatives explained that this was due to the fact that the contract was let at a much earlier time prior to issuance of Amendment 13 and the subsequent commitments.
Much of this equipment is currently at the site and is in warehouse storage.
The licensee is reviewing courses of action that need to be taken to ensure that this equipment will meet the IEEE 323-1974 requirements in accordance with their commitment.
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This matter is considered to be unresolved until it has been determined whether or not this NSSS has been properly qualified in accordance with the licensee's commitment, and exactly what equipment has or has not been qualified.
12.
FSAR Docketir.g The FSAR (Final Safety Analysis Report) was docketed by the NRC on August 25, 1981, and will be considered the governing document involving licensee commitments for future inspections.
However, Chapter 17.1 of the PSAR (Preliminary Safety Analysis Report) will still remain tha
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governing document relative to the licensee's construction QA program,
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in that this program is not referenced in the FSAR.
13.
Unresolved Items
Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or t
i deviations.
Four such items have been discussed within this report.
They will be entitled as follows in future discussions:
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Paragraph 8
"Reportability Requirements Relative to the Use of ASTM
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A-500 Grade B Materials" l
Paragraph 8
" Procurement Requirements Relative to the Use of ASIM A-500 Grade B Materials" Paragraph 9
" Verification of Seismic and Enviromental Qualification Documentation"
Paragraph 11
" Qualification of NSSS Equipment to IEEE by 323-1974 j
Requirements" 14.
Management Interviews The RRI met with one or more of the persons identified in paragraph 1 at various times during the inspection period.
An exit meeting was held on September 1, 1981, to discuss various findings and observations made during this inspection period.
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