ML20058B110
| ML20058B110 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1982 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| References | |
| NUREG-0040, NUREG-0040-V06-N02, NUREG-40, NUREG-40-V6-N2, NUDOCS 8207220652 | |
| Download: ML20058B110 (178) | |
Text
_
NUREG-0040 Vol. 6, No. 2
.lCENSEE CONTRACTOR LND VEND 0R INSPECTION STATUS REPORT hUARTERLY REPORT fPRil 1982-JUNE 1982 h
,n
,NITED STATES NUCLEAR REGULATORY COMMISSION
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Ah) 7 g 2 O20731 040 R PDR
Available from N RC/GPO Sales Program Superintendent of Documents Government Printing Of fice Washington, D. C. 20402 A year's subscription consists of 4 issues for this publication.
Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 Microfiche of single copies are available from NRC/GPO Sales Program Washington, D. C. 20555 t
NUREG-0040 Vol. 6, No. 2 LICENSEE CONTRACTOR AND VEND 0R INSPECTION l STATUS REPORT i
l QUARTERLY REPORT APRIL 1982 - JUNE 1982
.___n.
2_ _ : = =
Manuscript Completed: June 1982 Date Published: July 1982 Prepared by: Region IV U.S. Nuclear Regulatory Commission Arlington, TX 76011 y "" ' 0%
b.
i
CONTENTS l
Page 1.
PREFACE.............................
iii 2.
REPORTING FORMAT (Sampl e ).......... vii 3.
CONTRACTORS WITH IE LETTERS AND....
ix SAMPLE LETTER....................
xi 4.
INSPECTION REPORTS.................
1 5.
INDEX OF INSPECTION REPORTS........
175 i
PREFACE A fundamental p emise of the Nuclear Regulatory Commission's (NRC) nuclear facility licensing and inspection program is that a licensee is responsible for the pre construction and safe operation of nuclear power plants.
The total pvernment-industry system for the inspection of nuclear facilities has been designed to provide for multiple levels of inspection and verification.
Licensees, contractors, and vendors each participate in a quality verification process in accordance with requirements prescribed by, or consistent with, NRC rules and regulations.
The NRC inspects to determine whether its requirements are being met by a licensee and his contractors, while the great bulk of the inspection activity is performed by the industry within the frame-work of sequential ongoing quality verification programs.
In implementing this multilayered approach, a licensee is responsible for developing a detailed quality assurance plan as part of his license applica-tion.
This plan includes the quality assurance programs of the licensee's contractors and vendors.
The NRC reviews the licensee's and contractor's quality assurance plans to determine that implementation of the proposed quality assurance program would be satisfactory and responsive to NRC regulations.
Firms designing nuclear steam supply systems, architect engineering firms doing design work on nuclear power plants, and certain selected vendors are currently inspected on a regular basis by the NRC.
NRC inspectors, during periodic inspections, ascertain through direct observation of selected activities (including review of processes and selected hardware, discussions with employees and selected record review) whether a licensee or contractor is satisfactorily implementing a quality assurance program.
If nonconformances with QA commitments are found, the inspected organization is requested to take appropriate corrective action and to institute preventive measures to preclude recurrence.
In addition to the QA program inspectioas, NRC also conducts reactive inspec-tions of the licensee's contractors and vendors.
These are special, limited scope inspections to verify that organizations supplying safety-related equipment or services to licensed facilities are exercising appropriate corrective / preventive measures when defects or conditions which could adversely affect the safe operation of such facilities are identified and that these organizations are complying with the NRC requirements which govern the evaluaticn and reporting of such conditions.
In the case of the principal licensee contractors, such as nuclear steam supply system designers and architect engineering firms, the NRC encourages submittal of a description of corporate-wide quality assurance programs for review and acceptance by the NRC Office of Nuclear Reactor Regulation (NRR).
Upon acceptance by NRR, described quality assurance programs provide written bases for inspection on a generic basis, rather than with respect to specific commitments made by a particular licensee.
Once accepted by NRR, a corporate quality assurance program of a licensee's contractor will be acceptable for all license applications that incorporate the program by reference in a Safety iii
}
Analysis Report.
In such cases, a contractor's quality assurance program will not be reviewed by NRR as part of the licensing review process, provided that the incorporation in the SAR is without change or modification.
However, new or revised regulations, Regulatory Guides, or Standard Review Plans affecting quality assurance program controls may be applied by NRR to previously accepted quality assurance programs.
The NRC Region IV Office In Arlington, Texas, inspects the implementation of quality assurance programs of nuclear steam supply system designers and architect engineering firms which have been submitted to, and approved by, NRR in the form of Topical Reports or Standardized Programs.
Upon completion of inspections confirming satisfactory implementation of QA programs, NRC will 1
issue a confirming letter to the nuclear steam system supplier or architect
)
engineering firm.
1 Licensees and applicants that have referenced the NRR approved Topical Report, or Standardized Program, in Safety Analysis Reports (or have adopted the total quality assurance program described in the Topical Report or Standardized Program) may, at their option, use the confirming letter to fulfill their obligation under 10 CFR Part 50, Appendix B, Criterion VII, that requires them to perform initial source evaluation audits and subsequent periodic audits to verify quality assurance program implementation.
For additional details concerning the NRC letter, refer to " SAMPLE LETTER" included in this report.
Licensees or construction permit holders may choose not to make use of a contractor's NRC accepted program, or such an accepted program may not exist.
In such cases, the Region IV inspections of nuclear steam supply system designers, architect engineering firms, or other licensee contractors, sub-tier contractors, or suppliers, will be based on programs developed to meet the commitments made by the licensee or construction permit holder.
These Region IV inspections will not relieve the licensees or applicants from any inspection / verification responsibilities required by Criterion VII.
The NRC currently is continuing their evaluation of a proposed program for NRC acceptance of third party (ASME) certification of Vendor Quality Assurance programs.
Should the proposed program be endorsed by NRC, it is anticipated that, subject to NRC audits of the third party program, licensees and applicants would be able to use the ASME nuclear certification and inspection system to fulfill that part of their obligation under 10 CFR Part 50, Appendix B, Criterion VII, which required them to perform initial source evaluation / selection audits and subsequent periodic audits to assess the quality assurance program implementation.
A third party category of firms consists of organizations whose quality assurance programs or manufacturing processes have not been reviewed and l
approved by NRR, or by a third party (such as ASME).
This category of firms is subject to NRC inspection based on the safety significance and performance of products or services provided by such firms.
Since such firms will not receive a third party review of their QA programs, results of the direct NRC inspections may not be used to fulfill the licensees's obligations under Criterion VII.
l iv t
The White Book contains information normally used to establish a " qualified suppliers" list; however, the information contained in this document is not
}
adequate, nor is it intended to stand by itself as a basis for qu'alification l
of suppliers.
1 Correspondence with contractors and vendars relative to the inspection data contained in the White Book is placed in the USNRC Public Document Room, located in Washington, D.C.
Copies of the White Book may be ob'ained at a nominal cost by writing to the National Technical Information Service, Springfield, Virginia 22161.
v
ORGANIZATION:
COMPANY, DIVISION CITY, STATE REPORT Doc ket/ Year INSPECTION INSPECTION NO.
Sequence DATF(S):
ON-SITF HOURS:
CORRESPONDENCE ADDRESS:
Corporate Name Division
" SAMPLE PAGE" ATTN:
Name/ Title Address City / State / Zip Code 1
ORGANIZATION CONTACT:
Name/ Title Telephone Number PRINCIPAL PRODUCT: Description of type of components, equipment, or services su ppl i ed.
NUCLEAR INDUSTRY ACTIVITY:
Briet statement of scope of activity including percentage of organization effort, if applicable.
ASSIGNED INSPECTOR:
(Signature)
Name/VPB Section Date OTHER INSPECTOR (S):
Name/VPB Section l
APPROVED BY:
(Signature)
Name/VPB Section Date i
INSPECTION BASES AND SCOPE:
A.
BASES: Pertain to the inspection criteria that are applicable to the activity being inspected, i.e.,10 CFR Part 21, Appendix B, and SAR or Topical Report
(
commitments.
B.
SCOPE: Summarizes the specific QA program areas that were reviewed, and/or l
identifies plant systems, equipment or specific components that were inspected.
l For reactive (identifici problem) inspections, the scope summarized the l
problem that caused the inspection to be performed.
PLANT SITE APPLICABILITY:
Lists docket numbers of licensed facilities for which equipment, services or records were examined during the inspection.
vii
ORGANIZATION:
COMPANY, DIVISION CITY, STATE i
REPORT NO-INSPFCTION RESULTS-Pa ce 2 o f 2 A.
VIOLATIONS:
Shown here are any inspection results determined to be in violation of Federal Regulations (such as 10 CFR Part 21) that are applicable to the organization being inspected.
B.
NONCONFORMANCES:
Shown here are any inspection results determined to be in nonconformance with applicable commitments to NRC requirements.
In addition to identifying the applicable NRC requirements, the specific industry codes and standards, company QA Manual sections, or operating procedures which are used to implement these commitments may be referenced.
C.
UNRESOLVED ITEMS:
Shown here are inspection results about which more information is required in order to determine whether they are acceptable items or whether a Violation or Nonconformance may exist.
Such items will be resolved during subsequent inspections.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
This section is used to identify the status of previously identified Violations, items of Nonconformance and/or Unresolved Items until they are closed by appropriate action.
For all such items, this section will indicate open/ closed classification, identify the inspection report which first produced the item, describe the nature of the item, and if closed, include a brief statement concerning action which closed the item.
If this section is omitted, all previous inspection findings have been closed.
E.
OTHER FINDINGS OR COMMENTS: This section is used to provide significant information concerning the inspection areas identified under " Inspection Scope."
Included are such items as mitigating circumstances concerning a Violation or Nonconformance, statements concerning the limitations or depth of inspection.
(Sample Size, type of review performed) and special circumstances or concerns identified for possible follow up.
For reactive inspections, this section will be used to summarize the disposition or status of the condition or event which caused the inspection to be performed.
SAMP.LE PAGE EXPLANATION OF FORMAT AND TERMINOLOGY viii
CONTRACTORS WITH NRC LETTERS CONFIRMING QA PROGRAM IMPLEMENTATION (SEE NEXT PAGE FOR EXAMPLE OF CONFIRMING LETTERS)
COURACTOR IOPICAL REPORT REVISION DATE OF NRC LETTER PABCOCK & WILCOX PAW 10096A REVISION 1 DECEMBER 30, 1975 STONE & WEBSTER SWSQAP1-74 REVISIm A DECEMBER 30, 1975 WESTINGHOUSE NTD WCAP-8370 REVISIm 9A APRIL 30,1981 BECtHEL-GAITHERSBURG BQ-TOP-1 REVISIm 2A MARCH 16, 1981 BECHTEL-SAN FRANCISCO BQ-TOP-1 REVISIm 3A JUNE 12,1981 EBASCOSERVICES,INC.
ETR-1001 PsvlSIm 8A MARCH 31, 1980 COMBuSTim ENGINEERING CENPD-210-A REVISIm 3 JUNE 2,1981 GIBBS & HILL, INC.
GIBSAR17-A AMENDMENT 2 MARCH 31, 1977 SqTED fjeGINEERS &
LONSTRUCTORS UEC-TR-001-3A AMENWENT 5 MARCH 31,1977 GENERAL ELECTRIC CO.
NED0-11209-04A N/A MARCH 30,1977 SgRGENT&LUNDY tNGINEERS SL-TR-1A REVISION 5 M4Y17,1979 BECFREL-LOS ANGELES GO-TOP-1 REVISIm 2A JUNE 5, 1979 i
GILBERT-C0oNONWEALTH GAI-TR-106 REVISIm 2A FEBRUARY 2,1981 BECHTEL-ANN ARBOR BO-TOP-1 REVISim 2A PAY 7, 1981 l
iX
UNITED STATES
- 'u a,}o o
NUCLEAR REGULATORY COMMISSION
/
REGION IV A*( ^ l
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611 RY AN PLAZA DRIVE SUITE 1000
- j ARLINGTON, TEX AS 76012
..... "/
- , 4 (ADDRESSEE)
Gentlemen:
A series of Nuclear Regulatory Commission (NRC) inspections have been conducted to review your implementation of the Quality Assurance Program applicable to NRC applicants or licensees who have contracted for services from the (applicable corporate entity).
These inspections consisted of selective examination of procedures and representative records, interview of personnel, and direct observation by the inspectors.
As a result of these inspections, the NRC has concluded that the QA program described in Topical Report is being implemented satisfactorily.
Neither this conclusion nor the remainder of this letter applies to manufacturing activities or construction related activities conducted at reactor sites.
Licensees and applicants that have referenced the above Topical Report in their Safety Analysis Reports (or have adopted the total quality assurance program described in that Topical Report) may, at their option, use this letter to fulfill their obligation under 10 CFR Part 50, Appendix B, Criterion VII, that requires them to perform initial source evaluation /
selection audits and subsequent periodic audits to assess the quality assurance program implementation.
The NRC expression of satisfaction with the implementation of your QA program does not assure that a specific product or service offered by you to your customer is of acceptable' quality, nor does it relieve the applicant or licensee from the general provision of Criterion VII which requires verifi-cation that purchased material, equipment, or services conform to the procure-ment documents.
It is recognized that in some cases this assurance can be made by the applicant or licensee without audits or inspections at your facility.
Continuing acceptability of implementation of your QA program is contingent upon your maintaining a satisfactory level of program implementation, certified through periodic NRC inspection, throughout all corporate organiza-tion units and nuclear projects encompassed by your program.
Should your program implementation at any time be found unacceptable you will be notified by letter and requested to correct the deficiencies promptly.
In the xi
(ADDRESSEE) (DATE) event you fail to correct the deficiencies promptly, or if the record of deficiencies is such as to indicate generally poor program implementation, you and the applicants and licensees who have referenced your QA program will be notified that the generic implementation of your program is no longer acceptable to the NRC.
All of the audit / inspection requirements of Criterion VII, Appendix B, 10 CFR Part 50, must then be implemented by the applicants or licensees.
The NRC will reinstate its letter of acceptability of implementa-tion of your QA program only after our inspectors have concluded, based on reinspection, that you have again demonstrated full compliance.
Except as noted above, the conclusions expressed in this letter will be effective for three years from the date of issue of the letter.
At that time, program performance over the previous three year period will be evaluated and this letter reissued, if appropriate.
The results of our inspection are published quarterly in the Licensee Contractor and Vendor Inspection Status Report (NUREG 0040), which is made available to NRC facility applicants, licensees, contractors, and vendors, as well as to members of the public, by subscription.
Sincerely, Regional Administrator xii
ORGANIZATION:
ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORI INSPECTION INSPECTION N0.
99900053/82-01 DATE(S) 2/23-26/82 ON-SITE HOURS: 23 CORRESPONDENCE ADDRESS: Anchor Darling Valve Company ATTN:
Mr. J. W. Marlatt General Manager 701 First Street Williamsport, PA 17701 ORGANIZATIONAL CONTACT:
Mr. G. W. Kneiser, Quality Assurance Manager TELEPHONE NUtiBER:
(717) 323-6121 PRINCIPAL PRODUCT: ASME Section III Class 1, 2, and 3 valves.
NUCLEAR INDUSTRY ACTIVITY: Anchor Darling Valve Company's contribution to the nuclear industry represents approximately 70 percent of its total work load.
i A
s' s 6!<91 ASSIGNED INSPECTOR:
+
j
/,
Kelley,Readt'ive&Cf.ponentProgramSection Da'te '
Wm. D.
(R& CPS)
OTHER INSPECTOR (S):
APPROVED BY:
M, M,+- e 9- / s-/&t~~
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21 and 10 CFR Part 50, Appendix B.
B.
SCOPE:
This inspection was made as a result of the issuance of: (1) a con-structica deficiency report by Mississippi Power and Light Company concerning the premature opening of torque switches on valves installed at the Grand Gulf Nuclear Plant which resulted from engagement by excessively threaded stem protectors with the stem lock nuts; and (2) a potential 10 CFR Part 21 report by Pennsylvania Power and Light Company (PP&LC) concerning cavitation (cont. on next page) f PLANT SITE APPLICABILITY:
50-387, 50-416
_________ _______________ ____ a
ORGANIZATION:
ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INdetLilun NO..
99900053/82-01 RESULTS:
PAGE 2 of 4 SCOPE: (Cont.) and/or dif ficulty in the throttling of valves F017A and B within the ranges specified in the PP&LC Technical Specification for the start-up testing of the Susquehanna Steam Electric Station's Residual Heat Removal System.
Additional areas selected for inspection included design and document control, and status of previous inspection findings.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None D.
STATUS O! PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Unresolved Item (81-01) - The inspector verified by review of the manufacturing drawings for the original valves installed in the Southern California Edison Company's San Onofre Nuclear Generating Station that the stacked tolerances were controlled during manu-facturing and assembly but were not in accordance with the clearances specified in the ADVC Design / Drafting Standard for quick closing valves.
The valves installed in the San Onofre Nuclear Generating Station were modified at the licensee's request to increase the closing speed and the ADVC standard was not developed until subsequent to this modifi-l cation.
2.
(Closed) Nonconformance (81-01) - Two gate valves on Order No.
1024-001-001 for Washington Public Power Supply System Nuclear Project 2 were shipped with the by pass installed on the opposite side of the valves.
The inspector verified that the ADVC quality assurance program now requires separate travelers for valves that are identical with the exception of the by pass.
3.
(0 pen) Nonconformance (81-01) - An equation for a circular plate with inner edge fixed and supported and uniform loading along the outer edge was improperly used for analysis of a flange.
This item was not reviewed during this inspection as a result of corrective action correspondence being incomplete.
2 l
ORGANIZATION:
ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPEC110N NO.
99900053/82-01 RESULTS:
PAGE 3 of 4 E.
OTHER FINDINGS OR COMMENTS:
1.
10 CFR Part 21 Report by Pennsylvania Power and Light Company (PP&LC) -
Problem reported was cavitation and/or difficulty in the throttling of valves F017A and B within the ranges speci.ied in the PP&LC Technical Specifications for the start up testing of the Susquehanna Steam Electric Station's Residual Heat Removal System.
Valves F017A and B were purchased from ADVC by Bechtel Power Corpora-tion (BPC) on their purchase orders 8856P-10-A and 8856-12-A, dated December 26, 1973.
The flow requirement for the valves was specified in the Valve Data Sheets which were attached to the BPC valve speci-fications 8856P-10 & 12; however, the specifications did not require flow tests be performed on the valves.
ADVC's selection of their valves to meet the required flow rates was based on their calculated C factors.
y BPC's letter to ADVC dated May 6, 1981, listed the throttled flow condition for six valves during plant start-up and the duration (hours / year) of the throttled conditions.
These throttled flows were not specified in the original purchase documents.
ADVC evaluated the new flow conditions and advised BPC that two of the volves (Item 9.9 on purchase order number 8856P-12-A and item 7.11 on purchase order number 8856P-10-A) would require modification of the discs.
Subsequently, BPC issued Revision 24, dated November 11, 1981, to purchase order number 8856-10-A and Revision 29, dated December 4,1980, to purchase order number 8856-12-A purchasing replacement throttling discs for the valves.
2.
Potential Construction Deficiency Report by Mississippi Power and l
Light Company (MP&LC) - Problem reported was the excessive threads l
on the valve stem protectors permitted the protectors to come in contact with the valve operator stem lock nut of the valves installed at the Grand Gulf Nuclear Plant, causing the valve operator opening torque switches to trip prematurely.
The Potential Construction Deficiency Report made by MP&LC to NRC Region II on March 3, 1980, identified two manufacturers whose stem protectors came in contact with the valve operator stem lock nuts, but did not identify the valves or the size and type of the Limitorque Corp. (LC) valve operators.
3
ORGANIZATION:
ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA KLPURI IN6f t.L e IUH NO..
99900053/82-01 RESULTS:
PAGE 4 of 4 The NRC inspector verified by visual observation of the LC valve operators (Models SMB-0, SMB-00, SMB-000, SMB-1, and SMB-2) and the LC catalog, that the valve operators had a thread run-out approximately 3/8 inch deep and a shoulder at the bottom of the run-out which would have prevented the pipe nipple or pipe plug (used as a stem thread protector or closure) from engaging the stem lock nut of the valve operators.
The LC valve operator Model SMB-3 has a lh inch space from the bottom pipe thread to the lock nut.
This would mean that the pipe nipple used as a stem protector would have to have a running pipe thread in excess of 1 inches in length to engage the lock nut.
All of the LC valve operators inspected had their hand wheels mounted on the side of the operators, and there were no LC SMB-000 valve operators with the hand wheel mounted on the top of the operator available for inspection.
3.
Design and Document Control - Reviewed one section of the quality assurance manual, eight engineering standards, two customer specifi-cations, one customer purchase order, one master valve list, one material drawing list, one shop order instruction, one engineering data sheet, two valve assembly drawings, one design report, one design / drafting standard, and one flange design project.
Verified l
procedures had been implemented to control review and approval of design calculations.
4
ORGANIZATION:
BABC0CK AND WILCOX COMPANY NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORf INSPECTIGN INSPECTION NO.-
99900400/82-01 DATE(S) 1/26-28/82 ON-SITE HOURS: 42 CORRESPONDENCE ADDRESS:
The Babcock and Wilcox Company Nuclear Power Generation Division ATTN:
Mr. D. E. Guilbert, Vice Pres. & Gen. Mgr.
P. O. Box 1260 Lynchburg, VA 24505 ORGANIZATIONAL CONTACT:
Mr. E. L. Davis, QA Manager TELEPHONE NUMBER:
(804) 385-2895 PRINCIPAL PRODUCT: Nuclear Steam Supply Systems, engineering services and operating plant support.
NUCLEAR INDUSTRY ACTIVITY: The total effort committed to domestic nuclear activities is approximately 88% of the 1,300 employees of the Nuclear Power Generation Division.
Principal activities include the design and procurement of nine projects; Bellefonte, Midland, Washington Public Power Supply System, North Anna, and Pebble Springs, and providing engineering services under 129 service contracts and 38 fuel reload contracts.
ASSIGNED INSPECTOR:
U
's
.ogT-4- { - 22.
~
gl-D. F. Fox, Reactor; Systems Section (RSS)
Date OTHER INSPECTOR (S): W. D. Kelley, Reactive & Component Program Section I
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( (i.,
b 6J-d p 4 @ Z.
APPROVED BY:
(4 C. J. Hale, Chief, RS$
Date INSPECTION BASES AND SCOPE:
f0CFRPart50.AppendixB,TopicalReportBAW-10096AandthePSAR A.
BASES:
for the 2itainia Electric Power Company's North Anna 3 station.
B.
SCOPE:
Th,s inspection was made as a result of:
(1) a Babcock & Wilcox Company (B&W) 10 CFR Part 21 report that flow control valves, used in reactor coolant makeup and purification systems, were furnished to the Washington Public Powe' Supply System (WNP 1/4) and to the Portland General Electric Company (Pebble Springs) without provisions for assuring that the flow (Cont. on next page)
PLANT SITE APPLICABILITY:
Documentation / records identified with the following nuclear facilities were examined during this inspection:
Washington Public Power Supply System, Units 1 and 4, 50-460/513; Portland General Electric Company, Pebble Springs, 50-514/515; Virginia Electric Pcwer Company, North Anna 3, 50-404.
5
ORGANIZATION:
BABC0CK AND WILC0X COMPANY NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECI10N NO.-
99900400/82-01 RESULTS:
PAGE 2 of 7 SCOPE:
(Cont.) through the valves in their closed position meets minimum flow requirements; (2) a Virginia Electric Power Company (VEPCO) 10 CFR Part 21/10 CFR Part 50.55(e) report that apparent disparities between the B&W analytical model of the VEPC0 North Anna 3 steam generator upper support system and the actual hardware design yielded loading data which could result in potentially nonconservative results when used by Stone and Webster to design the supporting steel and concrete structures; (3) valida-tion of computer codes; and (4) status of previous inspection findings.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
1.
Certain structural analyses may not meet regulatory requirements in that they do not appear to be sufficiently detailed with respect to I
assumptions, bases, source of inputs, reference to the hardware design drawings, analytical model-to-hardware relationship, and interpreta-tion of results.
This item will be further evaluated during a future inspection.
2.
Procurement controls for flow control valves do not appear to comply
/
with QA program commitments, in that valves were furnished that were not designcd to provide the required minimum flow of two gallons per minute in the closed position at a pressure differential of 820 pounds per square inch, and which B&W-NPGD source surveillance and vendor drawing review fuiled to detect.
This item will be further evaluated during a future inspection in regard to additional informa-tion that has been stated by B&W to be available for review at NPGD.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
(0 pen) Unresolved Item (81-02):
The recently appointed QA Manager does not appear to possess the minimum qualifications, for the position, that are delineated in the NRC approved revision of the B&W Topical Report.
B&W Topical Report BAW-10096 was revised (to draft rev. 4) and submitted to NRC headquarters for review and approval.
Resolution of this item was deferred pending NRC action on the proposed revision.
ll 6
ORGANIZATION:
BABC0CK AND WILCOX COMPANY NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPEC11UN l
NO..
99900400/82-01 RESULTS:
PAGE 3 of 7 E.
OTHER FINDINGS OR COMMENTS:
1.
Design Inspection (Validation of Computer Codes) - The B&W Topical Report (BAW-10096A), the NPGD (Nuclear) Quality Assurance Manual (19A-N.1) and implementing engineering and administrative procedures were reviewed to determine the commitments and requirements relative to control of analog and digital computer codes used in the design or analysis of safety related structures, systems and components.
Program (user's) manuals, code certification data and records, code usage and Control Data Corporation (CDC) 7600 central processing unit (CPU) run time data, technical Topical Reports, and other documentation related to:
(1) versions 13.0 and 15.0 of STALUM (a digital computer code used for the static and dynamic three-dimensional structural analyses of piping or structural support systems by a finite element method, all current versions of which were run approximately 4,200 times last year for a total of 118 CPU hours); (2) versions 13 and 17 of CRAFT 2 (a digital computer code used for the simulation and transient analysis of the transient multinode flows and momentums of a reactor plant during a loss of coolant accident, all current versions of which were run approximately 1,300 times last year for a total of 118 CPU hours); and (3) versions 7.4 and 8.4 of PDQ07 ( a digital computer code used to solve the reactor core neutron flux diffusion - depletion equations in three dimensions for up to five neutron energy groups, all current versions of which were run approximately 4,500 times last year for a total of 223 CPU hours).
The inspector noted that Revisions N and Q of the program manual for the STALUM computer code (NPGD-TM-376) were not revised in accordance with procedure NPGD-0903-03 in that the Technology Unit Manager responsible for reviewing the revised program manual did not sign in the appropriate block on the Title Page to attest that he reviewed the revised program manual for content and correctness as required.
Furthermore, the signature of the Technology Unit Manager on the Title Page of both revisions corresponded to an individual who terminated his employement at NPGD approximately one year before the revisions were issued.
7
ORGANIZATION:
BABC0CK AND WILC0X COMPANY NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORl INSPLLilON NO..
99900400/82-01 RESULTS:
PAGE 4 of 7 With respect to this item; i.e., reviewing and approving program manuals, two of the three organization units responsible for 1
generating program manuals did comply with procedure NPG-0903-03 in that the Technology Unit Manager responsible for reviewing the revised program manual did sign in the appropriate block on the Title Page of the revised manual.
l With respect to the actual methodology of verifying and certifying computer codes, and revisions thereto, the inspector noted the B&W engineering judgment that "the output of the calculation is reasonable compared to the input" even though documented, may not, depending on the test run, be a sufficient basis for certifying that the code will execute and produce results as described in the program manual for problems which more fully exercise the Code and the computer's temporary and permanent read / write data storage media.
This area will be inspected further during subsequent inspections.
2.
10 CFR Part 21/10 CFR Part 50.55(e) - This inspection was to determine the cause, significance, and design corrective action taken by B&W with respect to VEPC0 10 CFR Part 21/10 CFR Part 50.55(e) report that apparent disparities exist between the B&W analytical model, and the actual hardware design, of the upper supports for the VEPC0 North Anna 3 (NA-3) steam generators (SGs).
The reported safety concern was j
that the loadings on the supports under LOCA and other accident con-ditions that were calculated by B&W were nonconservative and could potentially yield nonconservative results when used by Stone and Webster (S&W) to design the embedment structures for the SG supports.
Review of available documentation and interviews with cognizant personnel indicate that B&W did calculate loading data, and transmitted it to S&W via Specification 18-1235000017-09 dated November 8, 1978.
The specification contained incongruent and nonconservative local maximum loading data with respect to upper SG support points 4 through 8 of the NA-3 SGs.
Specifically, (1) alternating loads (compressive and tensile loads in the Z-axis direction) were identified as being applied to SG upper support point 4 under accident conditions when the physical design of the support could not restrain tensile loads, and (2) the alternating loads (with respect to the Z-axis direction) that were identified as being applied to SG upper support points 5 through 8 were each only about one-half of what they should have been.
8
ORGANIZATION:
BABC0CK AND WILC0X COMPANY NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORI lh5PtLi1UN NO.
99900400/82-01 RESULTS:
PAGE 5 of 7 The incongruent and nonconservative local loading data apparently resulted from calculations made prior to November, 1978, which deployed an analytical model and structural analysis computer code (STALUM) with limited capabilities (i.e., STALUM could not accommodate boundary conditions (restraint points) that were not bidirectional),
and assumptions made by the analyst with respect to the mechanism for transmitting loads on the SG upper supports to their embedment structures.
Although not well documented in the calculations, it appears that the analyst assumed that any load in the Z-axis direction applied to either SG trunnion support beam was transmitted to the concrete embedment for each beam equally through the friction connections provided by pretensioned bolting at each end of each beam rather than totally to the appropriate shear plate povided at each end of each beam.
S&W requested B&W on November 4, 1980, to advise them whether the disparity between the mathematical model and the hardware design may have affected any data that had been transmitted to them.
B&W concluded that the data were affected by the disparity, and during a series of meetings and discussions held with S&W from October, 1980, through October,1981, both S&W and B&W agreed that:
(1) the loads applied at point 4 were not reported properly; and (2) the assumption that the loads applied to the concrete embedments through friction connections provided by pretensioned bolting was not viable due to bolt stress corrosion cracking concerns and the need for performing otherwise nonscheduled routine inspection and maintenance to assure adequate bolting pretension during plant operation.
B&W agreed to recalculate the loadings on the upper SG supports, and all other NSSS system component supports, where pretensioned bolting was assumed to provide friction connections for load transference to the support embedments.
The new calculations were also to include the effects of recently revised seismic response spectra, containment radial pressure pulse transient loadings, and other upgraded / revised loadings and displacements on NSSS components during accident conditions.
The calculations are being redone on a priority basis established by S&W sketch 81 ENG 6-1 dated November 30, 1981, and are being reported, as completed, via revisions to document 86-1127658.
All calculations are scheduled for completion by January 17, 1983.
B&W engineering stated that they planned to revise specification 18-1235000017 (Reactor Coolant System Foundation and Nozzle loadings) shortly thereafter.
9
ORGANIZATION:
BABC0CK AND WILCOX NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORI IN5PLtTION NO.
99900400/82-01 RESULTS:
PAGE 6 of 7 The safety significance of this design / analysis deficiency could not be established at B&W since the detailed design and the load bearing capabilities of the shear plates could only be ascertained at S&W.
However, since the data is being revised / upgraded prior to the final design and construction of the concrete embedments (as stated by B&W), there should be no impact on plant safety with respect to this item.
One unresolved item was identified in this area of the inspection (see C.1 above).
3.
Flow Control Valves in the Makeup and Purification System - B&W-NPGD notified the Nuclear Regulatory Commission (NRC) by letter dated January 9, 1981, of the failure of the 2 x 2 " 1,500 pound globe valves to meet the two gallon per minute flow requirement at a pressure differential of 820 pounds per square inch when the valves were in the closed position as specified in the B&W-NPGD Design Specification 08-1148000002-05.
These valves were supplied to Washington Public Power Supply System (WPPSS), Nuclear Projects Units 1 and 4 and Portland General Electric (PGE), Pebble Springs Nuclear Plant for use as makeup flow control valves in the makeup and purification systems.
B&W-NPGD notified WPPSS and PGE of the defiency ar.d initiated a program to determine a method of providing the required flow when the valves are in the closed position.
The valves for each project are identified as B&W Mark Number MU-V46.
An NRC inspection was performed at B&W Control Components incor-porated (B&W-CCI), Irvine, California, on June 22-24, 1981, to ascertain the cause of the failure of the valves to provide a two gallon per minute flow rate at a pressure differential of 820 pounds per square inch in the valve closed position.
As a result of this inspection (see NRC Report 99900262/81-01),
an NRC:RIV decision was made to follow up on the reported defi-ciency at B&W-NPGD.
10
ORGANIZATION:
BABCOCK AND WILC0X COMPANY NUCLEAR POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORi INSPECTION NO.
99900400/82-01 RESULTS:
PAGE 7 of 7 During this inspection the NRC inspector reviewed the B&W-NPGD Quality Assurance Manual, two change orders to two B&W-NPGD purchase orders, two change order acknowledgments (B&W Form BWNP-20089(1-76)), two inspection forms, one B&W-CCI test procedure, and one B&W-CCI drawing.
It was verified that: (1) B&W-NPGC had notified B&W-CCI, by change notices to their purchase orders, of the two gallon per minute flow requirement at a pressure differential of 820 pounds per square inch when the valves were in the closed position; and (2) B&W-CCI had acknowledged the change notices to the purchase orders that specified the flow requirement with the valves in the closed position.
With respect to the two gallon per minute minimum fl'ow requirement, the NRC inspector identified that: (1) flow tests had not been required by the B&W-NPGD Equipment Specification to assure compliance with the requirements; (2) cn.ece surveillance activity failed to detect that no provisions had been made by B&W-CCI in the valve design for compliance with this minimum flow requirement; (3) the B&W-CCI test procedure did not include a flow test to demonstrate compliance; (4) B&W-CCI drawing 920706018 was approved by B&W-NPGD, although it did not include this flow requirement; and (5) the valve deficiency had not been documented as a nonconformance and corrective action had not been documented or implemented to l
preclude recurrence.
B&W personnel stated at the exit meeting, that additional information was available on this subject that the NRC inspector did not have the opportunity to review. This item is considered unresolved pending review of this additional information.
11
ORGANIZATION:
BECHTEL POWER CORPORATION GAITHERSBURG POWER DIVISION GAITHERSBURG, MARYLAND REPORT INSPECTION 1/25-29/82 INSPECTION NO 99900519/82-01 DATE(5) 2/22-26/82 ON-SITE HOURS: 63 CORRESPONDENCE ADDRESS:
Bechtel Power Corporation Gaithersburg Power Division ATTN:
Mr. J. M. Komes, Vice President and General Manager 15740 Shady Grove Road Gaithersburg, MD 20760 ORGANIZATIONAL CONTACT:
Mr. D. C. Kansal, QA Manager TELEPHONE NUMBER:
(301) 258-3776 PRINCIPAL PRODUCT: Architect Engineering Services NUCLEAR INDUSTRY ACTIVITY: The Gaithersburg Power Division has a total of 2,755 employees of which 2,017 or 73% are assigned to nuclear projects.
Major projects include Callaway Unit 1, Wolf Creek Unit 1, and Grand Gulf Units 1 and 2.
There are also modification / repair / service contracts on 14 additional recctor units.
ASSIGNED INSPECTOR:
- h. k. (hM, 71 J
. Costello, Reactor Systems Section (RSS)
O e OTHER INSPECTOR (S): W. F. Ang, Region II APPROVED BY:
~/7. /V. 3
%/f;Z.
C. J. Hale, Chie RSS Dale ~
INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B, and SNUPPS PSAR, Chapter 17 B.
SCOPE:
Implementation inspection of Bechtel's commitments in t:ie areas of compliance with NRC IE Bulletins 79-14 and 79-02, design corrective action, and a 10 CFR Part 50.5b(e) report and a potential construction deficiency report which covered the following:
(1) valve chatter in RHR heat exchanger vent valves (Grand Gulf 1 and 2); and (2) incorrect sized cable which resulted in a failure of the control rods to scram (Grand Gulf 1 and 2).
PLANT SITE APPLICABILITY:
The contents of this report relate to the following dockets: 50-416, 50-482, 50-483, 50-321, 50-366, 50-250, 50-251, 50-348, and 50-364.
~
13
ORGANIZATION:
BECHTEL POWER CORPORATION GAITHERSBURG POWER DIVISION GAITHERSBURG, MARYLAND
-REPORI INSFtLil0N NO.-
99900519/82-01 RESULTS:
PAGE 2 of 5 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None D.
OTHER FINDINGS OR COMMENTS:
1.
Design Corrective Action - SNUPPS PSAR Chapter 17, Engineering Depart-ment and Engineering Project Procedures were examined to determine quality assurance program commitments.
To verify implementation of these commitments, the following documents were examined: QA program management corrective act'on log, project open items list, three management corrective action reports, five open items, six nonconfor-mance reports (NCRs), four deficiency / noncompliance reports, three revised drawings, and three revised calculations.
Relative to the documents examined, all quality assurance program commitments were being properly implemented.
2.
Potential Construction Deficiency Report - Valve chatter in RHR Heat Exchanger Vent Valves (Grand Gulf).
Mississippi Power and Light Company (MP&L) reported to Region II that the RHR heat exchanger vent valves, manufactured by Yarway, open and close intermittently due to a worm gear in the valve operator which is not self-locking.
The purpose of this inspection was to determine if this was a generic prob;em, and if there was a breakdown in the l
AE's quality program.
The problem was discovered by Bechtel during testing operations at the construction site prior to turnover to MP&L and was, therefore,
(
not reportable under 10 CFR Part 21, but was rcportable under 10 CFR Part 50.55(e).
The problem was corrected by rewiring the control circuit to actuate the position switch which opens at full close or full travel of the valve and thus prevents reenergizing the circuit.
This position l
switch is in the switch compartment furnished with the Limitorque actuator and is available for customer use, if necessary.
- Normally, 14 i
ORGANIZATION:
BECHTEL POWER CORPORATION GAITHERSBURG POWER DIVISION GAITHERSBURG, MARYLAND HLPORI inartcitua NO.
99900519/82-01 RESULTS:
PAGE 3 of 5 the Limitorque valve actuator has a self-locking capability and it is not necessary to use the position switch lockout.
- However, in a few cases it is not possible to achieve self-locking capability and at the same time satisfy specification requirements of quick close operation at reduced voltage and high differential pressure.
This is what happened at Grand Gulf and affected 20 valve actuators out of a total of approximately 160.
The condition is difficult to predict prior to final test operation, but can be corrected by rewiring the control circuit on those valve actuators to achieve the self-locking capability.
No similar problems were found in other nuclear plants being designed by Bechtel/Gaithersburg.
However, a management information bulletin was sent cut to other Bechtel offices to alert them of the problem.
The potential error in this case appears to be a failure of the valve supplier to alert the user that this condition could occur and could be corrected by rewiring the control circuit.
All modifications have been completed on Unit 1.
Unit 2 modifications will be completed prior to Unit 2 fuel load.
No nonconformances or unresolved items were identified in this area of the inspection.
3.
Petential Construction Deficiency - Incorrect Sized Cable Resulting in Failure of the Control Rods to Scram On August 26, 1981, Mississippi Power & Light Company notified Region II that during control rod testing, 14 scram pilot valves were found stuck in the energized state when both of the solenoids were deenergized.
The purpose of this inspection was to determine if this was a generic problem and if there was a breakdown in the AE's quality program.
The valves that failed did not contain a defect.
They were damaged by being operated with insufficient voltage being supplied to the solenoid coils by the Reactor Protection System (RPS) due to cables of insufficient capacity.
The low voltage caused chattering of the solenoid core internals, resulting in damage and subsequent sticking of the internals, preventing proper operation.
G.E. Specification 22A3899AE CRD Hydraulic Control System, Section 4.3, states that the minimum pick up voltage for the pilot scram valves is 87 VAC.
Bechtel calculations showed the minimum voltage across the most remote solenoid valve to be 96 VAC with the original sized wiring.
15 1
ORGANIZATION:
BECHTEL POWER CORPORATION GAITHERSBURG POWER DIVISION GAITHERSBURG, MARYLAND REP 0HI INSPLClauN NO.
99900519/82-01 RESULTS:
PAGE 4 of 5 After the problem had been discovered, information supplied by the solenoid manufacturer indicated their solenoids would not operate properly below 105 VAC and that they had been purchased to operate between 105 and 125 VAC The affected Reactor Protection System Supply Cables are being replaced with cables of sufficient capacity and all scram pilot valves are being rebuilt to replace damaged parts.
Two Nuclear projects at Bechtel/Gaithersburg use GE nuclear steam supply systems.
This problem was identified as being unique to the Grand Gulf project and did not affect the Hatch project.
The error appears to be in the GE/Bechtel interface whereby the information supplied by the solenoid manufacturer of a minimum requirement of 105 VAC for proper operation was not transmitted in GE documentation to Bechtel.
A Problem Alert was issued by Bechtel/Gaithersburg to other Bechtel offices on January 6, 1982.
This problem alert warned of the danger of low voltage with this type solenoid.
No nonconformances or unresolved items were identified in this inspection.
However, in future inspections the NSSS/AE interface requirements will be examined in greater detail.
4.
Follow up on IE Bulletins 79-14 and 79 This was a follow-up inspection related to activities performed at the Gaithersburg office of Bechtel Power Corporation which resulted from the issuance of IE Bulletin 79-14 (Seismic Analysis fnr As-Built Safety Related Piping Systems) and IE Bulletin 79-02 (Pipe Support Base Plate Designs Using Concrete Expansien Anchor Bolts).
The inspection was performed by an inspector from Region II with limited participation from the Region IV inspector.
There were no findings regarding the programmatic aspects of Bechtel's ef forts.
The other aspects of this inspection will be reported in Region II Reports 50-416/82-06; 50-321/82-03; 50-366/82-03; 50-348/82-07; 50-364/82-06; 50-250/82-12; and 50-251/82-12.
The highlights of these inspections are as follows:
a.
Grand Gulf, Unit 1, Report No. 50-416/82 To verify imple-mentation of the requirements of IEB 79-02 and 79-14, six stress problems and six baseplate design calculations with their related documents were examined.
Also, the as-built configuration for two pipe anchors was checked.
One unresolved item was 16
ORGANIZATION:
BECHTEL POWER CORPORATION GAITHERSBURG POWER DIVISION GAITHERSBURG, MARYLAND REPORT A N5PtL 61Uti NO.
99900519/82-01 RESULTS:
PAGE 5 of 5 identified to the licensee.
There were no violations or devia-tions.
The unresolved item concerned IEB 79-02 and 79-14 analytical discrepancies of which six were identified.
b.
Hatch, Units 1 and 2, Report Nos. 50-321/82-03 and 50-366/82 To verify implementation of the requirements IEB 79-02 and 79-14, three Licensee Event Reports, seven drawings and one letter were examined.
One unresolved item was identified to the licensee.
There were no violations or deviations.
The unresolved item concerned three unresolved discrepancies involving IEB 79-02 and 79-14.
c.
Joseph M. Farley, Units 1 and 2, Report Nos. 50-348/82-07 and 50-364/02 To verify implementation of the requirements of IEB 79-02 and 79-14, one schematic drawing, one isometric drawing, and one stress problem with their related documentation were examined.
No violations, deviations or unresolved items were identified in the IE 79-02 and 79-14 area.
Ilowever, one unresolved item was identified to the licensee concerning the nonsafety-related classification of the diesel generator exhaust piping and pipe supports.
d.
Turkey Point, Units 3 and 4, Report Nos. 50-250/82-12 and 50-251/82 To verify implementation of the requirements of IEB 79-02 and 79-14, two stress problems with their related docu-mentation were examined.
One unresolved item concerning two discrepancies involving IEB 79-14 was identified to the licensee.
Also, another unrelated unresolved item was identified to the licensee concerning the nonsafety-related classification of the diesel generator exhaust piping and pipe supports.
No violations or deviations were identified.
17
ORGANIZATION:
BECHTEL POWER CORPORATION LOS ANGELES POWER DIVISION NORWALK, CALIFORNIA REPORI INSPECTION INSPECTION NO.
99900521/82-01 DATE(S) 3/8-12/82 & 3/22-26/8, ' ON-SITE HOURS: 64 CORRESPONDENCE ADDRESS:
Bechtel Power Corporation Los Angeles Power Division ATTN:
Mr. J. V. Morowski Vice President and General Manager P. O. Box 60600, Terminal Annex Los Angeles, CA 90060 ORGANIZATIONAL CONTACT:
Mr. R. L. Patterson, QA Manager TELEPHONE NUMBER:
(213) 864-6011, Ext. 2061 PRINCIPAL PRODUCT. Architect Engineering Services NUCLEAR INDUSTRY ACTIVITY: The Los Angeles Power Division of the Bechtel Power Corporation is the architect engineer for nine domestic reactor units.
Fifty percent of the total personnel (approximately 7,700) are assigned to activities in connection with these units.
At the present time, the Houston Area office has a staff of 457.
k. d,., t. b 5/24 A2 ASSIGNED INSPECTOR:
i J. R. Costello, Reactor Systems Section (RSS)
Date OTHER INSPECTOR (S):
.'26h,~)-.
(7 APPROVED BY:
C. J. Maj,' Chi ~~f, W Date '
e INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B, and fopical Report No. BQ-TOP-1 B.
SCOPE:
The inspection conducted at Bechtel's Norwalk and Houston facilities was to assess the effectiveness of Bechtel's program for conducting audits and training, to check on the status of previous inspection findings, and to evaluate the quality program implications of two items reported per 10 CFR Part 50.55(e) and 10 CFR Part 21.
(Cont. on next page)
PLANT SITE APPLICABILITY:
The contents of this report relate to the following dockets: 50-361, 50-362, 50-424, 50-425, 50-498, 50-499, 50-528, 50-530.
19
ORGANIZATION:
BECHTEL POWER CORPORATION LOS ANGELES POWER DIVISION NORWALK, CALIFORNIA REPORT INSPECTION NO.
99900521/82-01 RESULTS:
PAGE 2 of 6 SCOPE:
(Cont.)
One item concerned the design of the primary loop crossover leg pipe restraint (Alvin W. Vogtle 1 & 2) and the other the omission of locking devices on fasteners of certain mechanical shock arrestors (San Onofre 2 & 3,.
A.
VIOLATIONS:
i i
None.
B.
NONCONCORMANCES:
1.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and Quality Assurance Department Procedure (QADP) 5.3, no department / project /
construction mrnagers or their designees attended the pre and post-audit cenferences for the management audit of SONGS 2 and 3 engineering office.
Also, no department / project / construction manager or their designees attended the pre-audit conference for the management audit of the Vogtle design office.
2.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and QADP 5.3, the response to CAR-002 of the management audit of the Vogtle Design Office was 11 days beyond the 30-day requirement and no scheduled date for response had been established.
Further the manage-ment audit of division engineering had scheduled corrective action for CAR-001 by March 20, 1981, however, this item is still open almost a year later.
CAR-001 required the establishment and documenting of a system that would provide for implementing quality requirements consistent with ANSI N45.2 in engineering initiated technical service agreements and consultant agreements.
3.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and QADP 5.1, the Vogtle Audit Activity Report for January and February 1982 did not identify the personnel contacted during the pre-audit, audit, and post-audit activities.
20
ORGANIZATION:
BECHTEL POWER CORPORATION LOS ANGELES POWER DIVISION NORWALK, CALIFORNIA l
REPORI INSPELiiUN NO.
99900521/82-01 RESULTS PAGE 3 of 6 4.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and Palo Verde Quality Program Procedure No. 18.0, Palo Verde Project Audit Reports PVH 5/18-29, PVH 7/81-16 and PVH5/81-9 did not include persons contacted during pre-audit, audit and post-audit activities.
No written response is required to this nonconformance as cor-rective action was taken prior to the close of the inspection.
Palo Verde Quality Program Procedure No. 18.0 was revised to require only the names and titles of persons contacted during the audit activities.
C.
UNRESOLVED ITEMS:
1.
It is not apparent that there is an effective means of revision controi the Quality Assurance Training Manual (which is a controlled document).
The inspector could not be certain the procedures in the manual were the latest pro-cedures since the Table of Contents did not reference revisions.
2.
It is not apparent that the present method of verifying indoc-trination and training of Quality Assurance Engineers and project personnel is adequate.
The practice of using signatures, initials, and dates in the training records to show completion of indoctrination and training does not provide a measure of comprehension of the subject matter.
Comprehension for NDE personnel, for example, is provided by written examination and is documented.
D.
STATUS OF PREVIOUS INSPELTION FINDINGS:
(Closed) Nonconformance (81-02):
Contrary to Criterion III of 10 CFR Part 50. Appendix B, and Appendix 17A of the Alvin W. Vogtle Nuclear Power Plant (VNP) PSAR, Bechtel had revised Section 17 21
ORGANIZATION:
BECHTEL POWER CORPORATION, LOS ANGELES POWER DIVISION NORWALK, CALIFORNIA REPORT INSPECTION NO.
99900521/82-01 RESULTS:
PAGE 4 of 6 of the VNP Project Reference Manual to allow Georgia Power Company field operations to make changes tc
- proved engineering drawings, specifications, or other design documents without prior Bechtel Project Engineering approval.
Georgia Power Company requested and obtained approval from NRR to change their Quality Program commitments which will allow them to make field changes to approved drawings, specifications, or other design documents without prior design approval from Bechtel.
This action legitimizes the actions taken by Bechtel in the above non-conformance. The inspector has been informed by the Georgia Power Company (GPC) Project QA Manager that documents have been placed in the SAR file confirming this change.
The inspector has also been informed that special procedures have been instituted by GPC to control this type of field change which requires an after the fact review of the change by Bechtel.
E.
OTHER FINDINGS OR COMMENTS:
1.
Audits - Applicable quality assurance and project procedures were examined to determine quality assurance program commitments.
To verify implementation of these commitments, the following documents were examined:
one management audit report, four manage-ment quality assurance master audit plans, three quality assurance project audit reports, five quality assurance audit reports, and two audit activity reports.
Relative to the documents examined, the nonconformances in B above were identified.
t 2.
Training - Section 2 of the South Texas Project Quality Program Manual, Section 1.3 of the Bechtel Procurement Supplier Quality Manual, and applicable quality assurance and engineering procedures were examined to establish quality assurance program commitments.
To verify implementation of these commitments, the following docu-ments were examined:
quality training manual, seven training records of quality assurance engineers, five project quality program indoc-trination and training courses, and three certifications of Supplier Quality Representatives.
Relative to the documents examined two unresolved items were identified (see C.1 and C.2. above).
22
ORGANIZATION:
BECHTEL POWER CORPORATION LOS ANGELES POWER DIVISION NORWALK, CALIFORNIA REPORT INSPECTION NO..
99900521/82-01 RESULTS:
PAGE 5 of 6 A matter of concern was identified which pertains to the requirement of EDP 5.34, which states, "The Project Engineer shall be responsible for the formulation of the indoctrination and training program." At the present time, the South Texas Project has a l
Project Engineering Manager and three Project Engineers.
The l
Systems, NSSS, Licensing, and QA Project Engineer have been responsible for the Phase A indoctrination and training program, but the present project procedures do not clearly specify this responsibility.
The training program for Phase B has not yet been fully conceptualized and the inspector is concerned that this responsibility be identified, particularly if more than one project engineer should be involved.
3.
10 CFR Part 21 (Vogtle) - This design deficiency was reported by Bechtel to Region IV.
The deficiency was discovered as a result of a request from GPC Construction to relax alignment and contact bearing requirements, resulting in a review of the crossover leg restraint design calculation.
During the course of this review, a deficiency was identified in the assumptions used in modeling the transfer of loads from the crossover leg restraint to the slab and containment basemat.
A recheck of the original engineer's work in seven other calculations resulted in two more designs being corrected.
The original designs had been design verified and the fault would have to be traced to the assumptions of the original engineer, and the fact that the design reviewer accepted these assumptions as reasonable.
Corrective action was taken by adding stiffener plates, splice plates, modifying bolt material, and changing bolt dimensions to exclude bolt threads from shear planes.
A review of the remaining NSSS restraint and support designs did not show any additional deficiencies.
This design is considered unique to the Vogtle project, but other Bechtel divisions were notified of this problem.
No nonconformances or unresolved items concerning quality program commitments were identified.
23
ORGANIZATION:
BECHTEL POWER CORPORATION, LOS ANGELES POWER DIVISION NORWALK, CALIFORNIA I REPORT INSPECTION N0.
99900521/82-01 RESULTS PAGE 6 of 6 4.
Deficiency Report 10 CFR Part 50.55(e) and 10 CFR Part 21 (San Onofre) - This deficiency was originally reported as a 10 CFR Part 50.55(e) by Southern California Edison and was later reported as a 10 CFR Part 21 by Bechtel.
The problem was first iaentified by a Bechtel Supplier Quality Representative on a surveillance inspection of ITT-Grinnel.
The supplier had failed to install hardened washers under the bolt heads in violation of the ASME code.
During a field survey to determine the extent of the use of soft washers in ITT-Grinnel mechanical shock arrestors, it was determined that fasteners on some Pacific Scientific supplied mechanical shock arrestors had been installed without washers or any locking device in violation of code requirements.
During the review of this problem it was found that a field change request to Construction Specification CS-P207 was issued, and incorrectly approved, to allow field installation of both ITT-Grinnell and Pacific Scientific snubbers without washers.
The problem has been corrected at San Onofre and a bulletin has been issued to other Bechtel divisions advising them of this problem.
No nonconformances or unresolved items were identified concerning quality program commitments.
i l
24
ORGANIZATION:
BECHTEL POWER CORPORATION SAN FRANCISCO POWER DIVISION SAN FRANCISCO, CALIFORNIA REPORI INSPECTION INSPECTION NO..
99900522/82-01 DATE(S) 4/19-23/82 ON-SITE HOURS: 32 CORRESPONDENCE ADDRESS:
Bechtel Power Corporation San Francisco Power Division ATTN:
Mr. C. D. Statton, V. Pres. and Gen. Mgr.
P. O. Box 3965 San Francisco, CA 94119 ORGANIZATIONAL CONTACT:
Mr. C. W. Dick, QA Manager IELEPHONE NUMBER:
(415) 768-0790 PRINCIPAL PRODUCT: Architect Engineering Services a
NUCLEAR INDUSTRY ACTIVITY: The total effort committed to domestic nuclear activities is approximately 92% of the 6700 person staff of the San Francisco Power Division.
The division currently provides the pr;ncipal architect engineering services for five domestic' units:
Limerick 1 and 2; Susquehanna 1 and 2; and Hope Creek 1; has project management for templetion of Diablo Canyon 1 and 2; has 12 units undee a modification / repair / service-type contract plus an engineering evaluation contract with an NSSS supplier.
ASSIGNED INSPECTOR:
.t.
_cf s ;
J. R. Costello, Reactor Systems Section (RSS)
Date OTHER INSPECTOR (S):
APPROVED BY:
\\
dh/
-I/9[, 7%
C. J.(Fiple, Chief! R5S Date~/ '
INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B, and Topical Report No. BQ-TOP-1.
B.
SCOPE:
This inspection was conducted to assess the effectiveness of Bechtel's QA program in the area of Evaluation of Supplier Performance /
Verification Activities and to check on the status of previous inspection findings.
PLANT SITE APPLICABILITY:
The contents of this report relate to the following dockets:
50-354, 50-387, l
and 50-388.
i 25
ORGANIZATION:
BECHTEL POWER CORPORATION SAN FRANCISCO POWER DIVISION SAN FRANCISCO, CALIFORNIA REPORT INSPECTION NO.
99900522/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
1.
It is not apparent that the requirements of EDPI-4.64.1 for reviewing and dispositioning quality surveillance reports are functioning properly.
Hope Creek Quality Surveillance Report No. 96 required procurement supplier quality action by July 10, 1981.
The inspector was unable to obtain any evidence during this inspection that the required action had been completed.
2.
It is not apparent that the requirements for specificity concerning codes and standards contained in EDP-4.49 are being satisfied.
EDP-4.49 required that applicable ccdes, standards and regulatory requirements have the issue and addenda properly identified and that these requirements be met in the design.
The inspector examined four Hope Creek project specifications and found that specification 10855-A-075(Q) did not specify the issue and addenda of four out of five codes and standards listed while speci-fication 10855-J-111(Q) did not specify the issue of ANSI N45.2 in effect.
The inspector will review this matter further on the next e
inspection.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Unresolved Item (80-03):
The position of relying on IE l
Bulletin activities to evaluate the significance of a potential safety hazard, instead of evaluating per 10 CFR Part 21, has been submitted to NRC Headquarters for review.
NRC Headquarters reply to the above concern stated that once a director or responsible officer of a company subject to 10 CFR Part 21, obtains information reasonably indicating that a defect exists, he must inform the Commission unless he knows that the Commission has already been adequately informed.
26
ORGANIZATION:
BECHTEL POWER CORPORATION SAN FRANCISCO POWER DIVISION SAN FRANCISCO, CALIFORNIA RLPORI Ih5PLCi10N NO.
99900522/82-01 RESULTS:
PAGE 3 of 3 In the two cases discussed in the 80-03 report, both were reported under 10 CFR Part 50.55(e).
The cases involved pilot solenoid valves and pipe supports for the Susquehanna project.
Both cases were referred to the 10 CFR Part 21 Evaluation Committee who determined that these defects did not exist on any other active projects at San Francisco Power Division (SFPD).
The committee also decided not to report these defects under 10 CFR Part 21 for nonactive projects at SFPD because the Commission was aware of the problem and was taking correttive action on all plants through IE Bulletins79-01A, 79-01B, and 79-14.
The inspector reviewed the action taken at SFPD and the required action requested by the IE Bulletins and concluded that the 10 CFR Part 21 Evaluation Committee actions were proper for these cases.
2.
(Closed) Unresolved Item (81-03):
The effectiveness of EDP-4.65 (Design Deficiency Processing) in implementing Section 9 (Design Corrective Action) of ANSI N45.2.11 could not be demonstrated during this inspection.
This area will be examined further during a future inspection.
Section II, No. 2 of the Nuclear Quality Assurance Manual (NQAM) has been revised as of April 19, 1982, to require Project QA to be responsible for documenting recommended corrective actions to minimize recurrence of design errors and to be responsible for tracking to ensure completion of the action.
This revision to the NQAM requires a change to or cancellation of EDP-4.65 which had required that engineering track CAR's.
This change in responsibilities should provide a more effective method of implementing Section 9 of ANSI N45.2.11.
E.
OTHER FINDINGS OR COMMENTS:
i Evaluation of Supplier Performance / Verification Activities - Applicable quality assurance and project procedures were examined to determine quality assurance program commitments.
To verify implementation of these commit-ments, the following documents were examined:
three quality surveillance reports, one project field supplier quality report, two supplier quality assurance n.anuals, five material requisitions, four supplier quality program evaluation reports, seven supplier quality program audit reports, four supplier evaluation review reports and three supplier performance evaluation reports.
Relative to the documents examined, two unresolved items were identified.
(See section C above.)
27
ORGANIZATION:
BURNS AND ROE, INC0gP0 RATED ORADELL, NEW JERSEY REPORT INSPECIION 2/23/82 and INSPECTION NO.
99900503/82-01 DATE(S) 3/15-18/82 ON-SITE HOURS: 20 CORRESPONDENCE ADDRESS:
Burns and Roe, Incorporated ATTH:
Mr. T. A. Hendrickson Assistant to the President S50 Kinderkamack Road Oradel1, NJ 07649 ORGANIZATIONAL CONTACT:
Mr. W. P. Rausch, Director of QA TELEPHONE NUMBER:
(201) 265-9351 PRINCIPAL PRODUCT: Architect Engineering and Consulting Services NUCLEAR INDUSTRY ACTIVITY: The total effort committed to domestic nuclear activities is approximately 32% of the 2260 employee staff.
Major projects include the design of Washington Public Power Supply System, Unit 2, post accident conditions update for Three Mile Island, Unit 2, and design verification for Oyster Creek.
l'/k 3-ASSIGNED INSPECTOR:
TC N
D. F. Fef}ReactorSysyehsSection(RSS)
D' ate '
OTHER INSPECTOR (S):
APPROVED BY:
Tl
/
h s
D'a t e ' ~
C. Y,Wa]e', Chi ef,'RSS INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B, and the FSAR for the Washington Public Power Supply Systems Unit 2 (WNP-2).
B.
SCOPE:
Determine the status of committed corrective action and preventive measures for previous inspection findings (Woodbury, New Jersey, offices) and design inspection (Richland Washington, offices).
PLANT SITE APPLICABILITY:
WNP-2, Docket No. 50-397.
29
ORGANIZATION:
BURNS AND ROE, INCORPORATED ORADELL, NEW JERSEY REPORT INSPECIION NO.
99900503/82-01 RESULTS:
PAGE 2 of 6 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and revision 1 to B&R Project Instruction WNP-2-ED-003 dated December 22, 1981, the Project Engineering Manager (WNP-2 Project, Woodbury Office) did not obtain a copy of the latest issue of Title 10, Chapter 1, Part 21, for inclusion as Exhibit II in the project version of procedure ED-003 as evidenced by the fact that the issue of 10 CFR Part 21 that was included as Exhibit II in the current project version of the project procedure (WNP-2-ED-003), was dated August 1, 1980, whereas, the the current issue of 10 CFR Part 21 is December 18, 1981.
Corrective actions and preventive measures were taken before the end of this inspection and no further written response is necessary.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and revision 1 to B&R Project Instruction WNP-2-PM-006 dated September 11, 1980, the receipt and number control of certain key project documents did not j
ensure that all revisions are received by the designated individual as evidenced by the fact that revision 5 of Project Instruction WNP-2-ED-010 (Calculations), dated June 23, 1981, was not received by the designated individual (the WNP-2 Project, Richland Office, holder of control copy No. 2) whose controlled set of Project Instructions contained superseded revision 4 of the Project Instruction WNP-2-ED-010, dated June 30, 1980.
3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 an.' revision 5 of B&R Project Instruction WNP-2-ED-010, dated June 23, 14"0, design calculation 5.51.053 (Environmental Analysis Outside Containment),
sub-analyses of which were dated December 23, 1981, through January 22, 1982, was not identified with respect to status, nor checked, in accordance with procedural requirements.
Specific examples of this nonconformance are as follows:
a.
Neither the lead sheet of calculation 5.51.053 nor the lead sheet of supporting sub-analyses (Events No. 1, 3, 4, and 10 others) were identified as " preliminary" or " final" even though all of the required lead sheets were signed and dated by the responsible Group Supervisor of the WNP-2 Project, Richland Office.
30
ORGANIZATION:
BURNS AND R0E, INCORPORATED OPADELL, NEW JERSEY REP 0HI INSPtC110N NO.
99900503/82-01 RESULTS:
PAGE 3 of 6 b.
The responsible Group Supervisor did not assign another engineer or designer to check the calculations as evidenced by the fact that the signature of the checker on the lead sheets for certain sub-analyses (events 1 through 3 and 5 through 12) cf calculation 5.51.053 was that of the responsible Group Supervisor of the WNP-2 Project, Richland Office, who also signed as the approver of these analyses.
c.
Documented evidence did not exist that the assigned checker actually examined calculation 5.51.053 in accordance with section 2.4 of Standard P015105G1 as evidenced by the omission from the calculation of the actual temperature, pressure, and
" harsh environment" time duration date, and the reference source of these design bases / input data, for which the motor operators for containment isolation valves were environmentally qualified.
This data is needed to support the conclusion of a sub analyses (Event 5) that the " Motor operators for containment isolation valves are environmentally qualified to postulated environment."
C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Deviation A.1 (80-01):
Design control measures were not being applied to verify or check the adequacy of design.
B&R Project Instruction WNP-2-ED-001 was revised as committed to include the requirement that "the preparer of a drawing, or change thereto, may not perform the required review by his discipline."
2.
(Closed) Violation (80-02):
Failure to meet 10 CFR Part 21 posting requirements in facilities where nuclear safety related activities were being conducted.
B&R Project Procedure PM-014.1 was revised and issued as ED-003 on April 15, 1981, as committed.
A notice, containing the information required by paragraph 21.6 of 10 CFR Part 21 was included as Exhibit I to Project Procedure ED-003 (and Project Instruction WNP-2-ED-003) and was posted in all affected facilities.
The posting was verified by QA surveillance reports dated January and August 1981, and February 1982, as committed.
)
31
ORGANIZATION:
BURNS AND ROE, INCORPORATED ORADELL, NEW JERSEY REPORT IN5PECTION i
NO.
99900503/82-01 RESULTS:
PAGE 4 of 6 One nonconformance was identified in this area of the inspection (see B.1 above).
B&R committed to revise Project Instruction WNP-2-ED-003 to reflect the current issue of the 10 CFR Part 21 regulations by March 31, 1982, and to review the procedure semiannually to assure currency.
No written response to this nonconformance is, therefore, required.
3.
(Closed) Deviation A (80-02):
Corrective action identified in B&R letter of response, dated June 5, 1980, to deviation A of NRC inspection report 80-01 was not completed as committed.
B&R revised Project Instruction WNP-2-ED-001 as committed (see D.1 above).
The importance of assuring that all corrective and preventive actions to NRC inspecticn findings have been completed as committed was reemphasized to the Quality Assurance organization and the WNP-2 Project QA Manager at the April 1981, QA Directors Staff Meeting.
4.
(Closed) Unresolved Item (83-02):
It appeared that B&R failed to report to NRC, under 10 CFR Part 21, that safety related switchgear was installed at WNP-2 in a manner (using top entry rigid electrical conduit) for which the switchgear had not been seismically qualified.
Wylie Laboratories have subsequently performed seismic testing and ana-lytical evaluation of typical Westinghouse switchgear installed with top entry rigid electrical conduit. They concluded in their report (WR-81-21, dated June 26, 1981), that the structural integrity of the switchgear, acceptable stress levels, and equipment response spectra levels for installed switchgear are equal to, or less than the originally qualified (by Westinghouse) switchgear levels.
The Wylie report is to be incorporated into the NRC/NRR seismic qualification review process.
Further action, if any, on this item is pending the results of the NRC/NRR review.
32 l
l
ORGANIZATION:
BURNS AND R0E, INCORPORATED ORADELL, NEW JERSEY REPORI INSPtCiiUN N0.
99900503/82-01 RESULTS:
PAGE 5 of 6 E.
OTHER FINDINGS OR COMMENTS:
Design Inspection - Branch Technical Position Papers, sections of the WNP-2 FSAR pertaining to postulated pioing failures in fluid systems and accident analyses, and applicable design control procedures were reviewed to determine the commitments, design bases and design measures employed by B&R to assure that the WNP-2 nuclear power station can be safety shutdown in the event of the postulated rupture of any high or moderate energy line out-side of tne reactor containment.
Drawings, instructions, calculations, specifications, reports, letters and memoranda, and other documentation related to the direct effects (pipe whip, jet impingement, etc.) and indirect effects (compartment temperature, pressure, humidity, water level, etc. ) resulting from a postulated pipe rupture were examined to verif. implementation of commitments.
The inspector noted that B&R pipe whip and jet impingement studies appear to be congruent with the current NRC guidance.
Furthermore, B&R WNP-2 Project Engineering personnel stated that these studies are to be verified against as-built drawings and piping system "walkdowns" at the WNP-2 site.
However, the inspector did not verify that the direct effects of a break in all high and moderate energy lines are scheduled to be eval-uated because the complete list of all such postulated breaks was not available at the B&R Richland Office.
Compartment temperature and pressure analyses were performed by the Washington Public Power Supply System (WPPSS) and were not available for examination at the B&R Richland Office.
Therefore, the NRC inspector could not verify that pipe ruptures identifies by WPPSS as requiring analyses for their effect on safe shutdown capability would represent the bounding or limiting cases for all compartments containing essential equipment in the WNP-2 nuclear power plant.
Two nonconformances, for failure to adhere to procedural requirements were identified in this area of the inspection (see B.2 and B.3 above).
1 33
ORGANIZATION:
BURNS AND R0E, INCORPORATED ORADELL, NEW JERSEY REPORT INSPECTION NO.
99900503/82-01 RESULTS:
PAGE 6 of 6 With respect to nonconformance B.2, the NRC inspector noted that a member of the B&R Project Director's staff was aware, for at least two weeks, that the WNP-2 Project Office's controlled copy of Project Instruction WNP-2-ED-010 was a year out of date, but had not notified the cognizant individual in Project Administration, nor otherwise caused it to be replaced with the current issue until identified by the NRC inspector.
With respect to nonconformance B.3.c, the NRC inspector determined that certain temperature, pressure, and " harsh environment" time duration data, were obtained via undocumented telephone conversations with WPPSS, who has assumed the responsibility for environmental qualification of all essential equipment for the WNP-2 site.
Further, (1) the functionability of mechanical devices, such as valves, under the environmental cenditions to which they may be subjected in the event of a pipe rupture has not been specifically addressed to date; and (2) sources of design input and design assumptions were not documented in inprocess calculation 5.51.050, (Pipe Breaks Outside Containment), making review of the calculation and its sub-analysts difficult without recourse to the originator. NRC Region V inspectic report 50-397/82-06 provides additional information in this area, as well s, other onsite B&R activities.
34
ORGANIZATION:
C&D BATTERIES, DIVISION OF ELTRA CORPORATION PLYMOUTH MEETING, PENNSYLVANIA REPORT NO.: 99900765/82-01 INSPECT 10N DATE(S):
1/4-8/82 INSPECTION ON-SITE HOURS: 25 CORRESPONDENCE ADDRESS: C&D Batteries Division of Eltra Corporation ATIN:
Mr. George M. Moses President 3043 Walton Rd.
Plymouth Meeting, PA 19462 ORGANIZATION CONTACT:
Mr. T. P. Davis, Director Quality Assurance TELEPHONE:
(215) 828-9000 PRINCIPAL PRODUCT: Batteries, Battery Chargers, and Battery Racks NUCLEAR INDUSTRY ACTIVITY:
One percent of production devoted to nuclear products used by Electrical Utilities.
m n
ASSIGNED INSPECTOR:
I2. IN 2.
L. B. Pa'rker, Reactive Inspection Section (RIS)
Dats OTHER INSPECTOR (S):
APPROVED BY:
" "z o 2 -, 2_ - y 2__
I. Barnes, Chief, RIS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B; and 10 CFR Part 21.
B.
SCOPE:
This inspection was made as a result of:
(1) the issuance of a 10 CFR Part 21 report by Bechtel Power Corporation concerning defective welds in Palisades Nuclear Plant battery racks; (2) a Licensee Event Report by Pacific Gas and Electric Company's Diablo Canyon Power Plant Engineer reporting cracks in the cells of the station batteries; and (3) the issuance of a 10 CFR Part 21 report by Alabama Power Company's Farley Nuclear Plant concerning defective battery post seals.
35
C&D BATTERIES, DIVISION OF ELTRA CORPORATION ORGANIZATION:
PLYMOUTH MEETING, PENNSYLVANIA REPORT NO.: 99900765/82-01 INSPECTION RESULTS:
PAGE 2 of 3 A.
VIOLATION:
Contrary to paragraph 21.21 of 10 CFR Part 21, procedures had not been adopted to implement 10 CFR Part 21.
B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
None.
This was the initial inspection of C&D Batteries, Plymouth Meeting, Pennsylvania.
E.
OTHER FINDINGS OR COMMENTS:
1.
Regional Requests a.
A 10 CFR Part 21 report by Bechtel Power Corporation, Ann Arbor, Michigan, was made concerning battery rack defective welds.
During modification and installation of the battery racks at the Palisades Nuclear Plant Site, several defective welds were observed.
C&D, Bechtel, and Consumer Power personnel inspected the defective welds at the Palisades site on October 14-15, 1981.
On October 16, 1981, Bechtel personnel inspected the prototype rack at Plymouth Meeting, Pennsylvania, (Prototype rack had been seismically tested by Wyle Laboratory, Huntsville, Alabama on Wyle Job No. 43450, December 7, 1976.)
Subsequent to these inspections the following actions were taken in regard to the rack weld deficiencies:
(1) C&D weld procedure attachments 1, 2, & 3 were added to C&D Welding Specification Section "B" W-1B by Rev. 2, on October 18, 1981; (2) the defective racks have been replaced at Palisades; (3) 20 RD 660 Frames, 8 RD-924 Frames, and 40 RE-2089 Brackets were returned to the East Greenville, Pennsylvania, C&D plant by Bechtel on C&D Return Material Authorization 22377; and (4) the USNRC RI and seven customers were notified by C&D of the battery rack deficiencies.
(See Follow-up Item E.3.)
b.
A Licensee Event Report by the Diablo Canyon Plant Engineer was made relative to cracking observed in the cells of the station batteries.
A C&D field representative visually inspected 300 cells (5 batteries) at the Diablo Canyon Nuclear Plant.
Ten of 36
C&D BATTERIES, DIVISION OF ELTRA CORPORATION ORGANIZATION:
PLYMOUTH MEETING, PENNSYLVANIA REPORT NO.: 99900765/82-01 INSPECTION RESULTS:
PAGE 3 of 3 the cells were found to be cracked / crazed, and one active cell was determined to be seeping.
Two other cells with seepage had been removed and were stored in the Diablo Canyon site maintenance shop.
At Millstone Nuclear Generating Stations 1 and 2, 142 cells were visually inspected by C&D personnel, with 3 cells being found crazed and no dampness evident on the bottom of the cells.
Twenty percent of the total rib areas in 120 cells at the Vermont Yankee Nuclear Generating Station were also inspected by C&D personnel with this inspection not revealing any leakage, dampness, or crazing / cracking of the containers.
In a letter to the USNRC RI, C&D Batteries stated that as a result of the above nuclear site inspections, and field inspections of the condition over the past nine years in both utility and telephone sites, they did not believe the condition represented a " substantial safety hazard."
In May 1972, the plastic cell container (type purchased in 1969 for use at Diablo Canyon) was redesigned and the integral rib block was replaced by a floating rib.
Pacific Gas and Electric Company, P. O. 4R56705 (November 6, 1981), was placed with C&D Batteries to replace the integral rib block cell batteries with redesigned batteries.
USNRC RI and 20 C&D customers were notified of these deficiencies.
c.
A 10 CFR Part 21 report by Alabama Power Company's Farley Nuclear Plant was made concerning a defective post seal design that allowed electrolyte fumes to escape.
C&D corrected this deficiency in early 1973; the new post seal utilizes a pure lead bushing molded into a hard rubber casing which, when cemented in place, reduces the electrolyte fumes which escape.
Farley was the only nuclear site which used batteries with the original design, C&D replacement batteries utilizing the new seal design were shipped to Farley on December 26, 1980 (Alabama Power Company No. 68501).
l 2.
Implementation of 10 CFR Part 21 - Posting in accordance with 10 CFR Part 21 was accomplished at the Plymouth Meeting, Pennsylvania plant.
(See A above for violation on paragraph 21.21 notification requirements.)
3.
Follow Up Item -
The adequacy of welding controls for battery rack fabrication was not reviewed during this inspection.
This subject and responses from C&D customers in regard to battery rack deficiency notification will be examined in a subsequent inspection.
37
ORGANIZATION:
COMBUSTION ENGINEERING, INC.
POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORI INSPECTION INSPECTION NO.
99900401/82-01 DATE(S) 3/22-26/82 ON-SITE HOURS: 27 CORRESPONDENCE ADDRESS:
Combustion Engineering, Inc.
Power Systems Group ATIN:
Mr. M. R. Etheridge, Vice Pres., Gen. Services 1000 Prospect Hill Road Windsor, CT 06095 ORGANIZATIONAL CONTACT:
Mr. C. W. Hoffman, Director, Group QA TELEPHONE NUMBER:
(203) 688-1911 PRINCIPAL PRODUCT: Nuclear Steam Supply Systems NUCLEAR INDUSTRY ACTIVITY: The Power Systems Group Combustion Engineering (CE) has contracts for 22 of the domestic reactor units to date, of which 14 are in the design and construction phase.
In addition, they have modification / repair /
service contracts for 22 reactor units.
V//#/r;t ASSIGNED INSPECTOR: 77, /d TJ/
au Da'te '
R.H.Brickley,gbactorSystemsSection(RSS)
OTHER INSPECTOR (S):
APPROVED BY:
1 Yb Date C. J(}l hfe, Chief, RSS J
INSPECTION EASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and Topical Report CENPD-210-A.
B.
SCOPE:
Design inputs, design document control, and status of previous inspection findings.
PLANT SITE APPLICABILITY:
Not Identified.
39
ORGANIZATION:
COMBUSTION ENGINEERING, INC.
POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORT INSPECTION NO.
99900401/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
Contrary to Section 17.5 of CENPD-210-A; Section 2.2 o Quality Assurance of Design Procedure (QADP) 5.3 (Interface Control) Revision 1, dated i
May 1, 1980; and Section 6.4.2 of Plant Engineering QA Procedures (PEQAP) l 001 (Design Quality Assurance Procedure) Revision 0, dated October 1, 1974; the following specifications did not have the required documentary evidence of interface reviews on file:
Droject Specification No. 6473-PE-231, Revision 5, dated January 12, 1981, (QADP 5.3); General Specification No. 00000-PE-707, Revision 2, dated Februa ry 19, 1975, (PEQAP 001).
C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (81-02):
Quality Assurance of Design Procedures do not address implementation of Section 9 (Corrective Action) of ANSI N45.2.ll-1974.
1 Previously completed corrective actions were identified in Inspection Report No. 99900401/81-03, paragraph D.1.
The remaining item, documen-tation of required training, was audited by engineering QA personnel, deficiencies identified and corrected, and all affected personnel were verified as receiving the required training.
2.
(0 pen) Nonconformance (81-03) - Certain operations, designated on the Integrated Manufacturing Quality Plan as being mandatory hold points, were performed by the supplier without either the Power Systems Group representative being present or a written waiver being issued.
Combustion Engineering's response to this nonconformance is under review by the NRC Region IV staff.
40
ORGANIZATION:
COMBUSTION ENGINEERING, INC.
POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORI INSPLLl10N NO.
99900401/82-01 RESULTS:
PAGE 3 of 3 3.
(0 pen) Nonconformance (81-03) - Record checklist requirements were not implemented in documentation packages, e.g.,
required material certi-fication reviews, material test reports of filler material, radio-graphs, and qualification of welders and welding procedure specifica-tions.
Combustion Engineering's respanse to this nonconformance is under review by the NRC Region IV staff.
4.
(0 pen) Nonconformance (81-03) - No in process audits were performed during fabrication of Component Cooling Water Heat Exchangers in regard to the welding processes used.
Combustion Engineering's response to this nonconformance is under review by the NRC Region IV staff.
5.
(0 pen) Nonconformance (81-03) - Group Quality Control surveillance or record review has not verified completion or fulfillment of Code special process requirements by external suppliers.
Combustion Engineering's response to this nonconformance is under review by the NRC Region IV staff.
E.
OTHER FINDINGS OR COMMENTS:
1.
Design Inputs:
Applicable procedures from the QADP Manual were examined to determine that procedures have been prepared and approved, consistent with program commitments.
Five System 80, eight project, one general specifications, and associated reviews and input veri-fication documents were examined to determine that the design input procedures were properly and effectively implemented.
One noncon-formance was identified and is discussed in paragraph B above.
2.
Design Document Control:
Applicable procedures from the QADP manual were examined to determine that procedurcs have been prepared and approved consistent with program ccmmitments.
Four System 80 and six project specifications were examined for internal distribution; and two project specifications were examined for external distri-bution to determine that design document control procedures were being properly and effectively implemented.
In the area inspected, no nonconformances were identified.
41
ORGANIZATION:
COMSIP, INC.
CUSTOMLINE DIVISION LINDEN, NEW JERSEY RLPORI INSPECTION INSPECTION t
NO.
99900771/82-01 DATE(S) 3/22-26/82 ON-SITE HOURS: 30 CORRESPONDENCE ADDRESS:
Comsip, Inc.
Customline Division ATTN:
Mr. C. M. Ferdinand, Manager, Quality Assurance 1418 E. Linden Avenue, P. O. Box 152 Linden, NJ 07036 ORGANIZATIONAL CONTACT:
Mr. C. M. Ferdinand TELEPHONE NUMBER:
(201) 486-1272 l
PRINCIPAL PRODUCT: Control Panels NUCLEAR INDUSTRY ACTIVITY: Approximately 50% of the work is devoted to the commercial nuclear industry.
- bi np, h.3 ut b/ 9- [d ASSIGNED INSPECTOR
- _/
y/ Section (R& CPS)
. E. Ellershaw, React.:ve & Component Program Date OTHER INSPECTOR (S):
f APPROVED BY:
e
$'- 2 c' 4 2_
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
This inspection was made as a result of a request by NRC's Region III office to inspect specific control panel fabricators due to identified weld defects found in panels that have been furnished to the Byron Station, Units 1 and 2; Braidwood Station, Units 1 and 2; Midland Plant, Units 1 and 2; and Callaway Plant, Unit 1.
Comsip, Inc., has provided main control panels to Callaway Plant, Unit 1.
PLANT SITE APPLICABILITY:
50-454 and 50-455; 50-456 and 50-457; 50-329 and 50-330, and 50-483.
43
ORGANIZATION:
COMSIP, INC.
CUSTOMLINE DIVISION LINDEN, NEW JERSEY REPORT INSPECIIGN NO.
99900771/82-01 RESULTS:
PAGE 2 of 5 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
i 1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, American Welding Society (AWS) Code Dl.1 and QA Manual Procedure 9.0, paragraph 5.1, neither the procedure specification nor the qualifi-l cation record of the gas metal arc welding process in Specification No. 5875-MP-1 addressed progression to be used during vertical position welding, which is an essential variable.
Further, the specific value for gas flow rate, an essential variable, l
was not reported in the qualification.
1 l
i 2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and AWS D1.1-76 l
paragraph 5.10.3, although the gas metal arc welding (GMAW) process in l
Speci:ication No. 5875-MP-1 has been used for making the fillet welds in the nain control panels for Palo Verde Nuclear Generating Station, Unit 3, no T-test fillet welds and resulting macroetch test specimens were made.
3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and QA Manual Procedure 9.0, paragraph 3.1, it has not been assured that welding, a special process affecting quality and subcontracted by Comsip for all nuclear work, has been accomplished using qualified personnel, in accordance with applicable codes and specifications, as evidenced by
(
the following observations made by the NRC inspector during a visit to H. R. Henrich, Inc., the subcontractor performing the fabrication /
welding of the main control panels for Palo Verde Nuclear Generating Station, Unit 3:
a.
The welding power sources (ammeters and voltmeters) are rive cali-brated nor were there any tong meters available to assure accuracy of the equipment.
b.
The following AWS 01.1 essential variables were being violated while GMAW was being performed:
44
ORGANIZATION:
COMSIP, INC.
CUSTOMLINE DIVISION LINDEN, NEW JERSEY REPORT INSPECI10N NO..
99900771/82-01 RESULTS:
PAGE 3 of 5-(1) Gas Flow Rate - the procedure requires 12-20 CFH, but 28 CFH was being utilized, which is in excess of the 25% maximum permitted increase in flow rate.
(2) Voltage - the procedure required 24-28, but 20-21 was being used, which is in excess of the permitted change of i 7%
from the specified mean arc voltage.
c.
An individual who was not qualified as a welder had performed plug welding using the gas tungsten arc welding (GTAW) process.
Further, a GTAW procedure did not exist, and there was no filler metal identity.
4.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, QA Manual Procedure 10.1 and AWS D1.1, paragraphs 3.6 and 3.10, the NRC inspector observed the following with respect to welds in painted control panels for Palo Verde Nuclear Generating Staion, Unit 3, which had been accepted in receiving inspection reports dated October 30, and December 18, 1981, by the Quality Assurance Representative:
a.
Undercut in excess of 1/32" in depth.
b.
Weld wire remnants up to 2 " long.
5.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and AWS D1.1, Comsip, Inc., reviewed and accepted a welding material certification from Metalfab, Inc., which did not provide chemical analysis of all of the elements required by the applicable AWS specification.
The certification states the welding material to be E-7014, which is included in AWS specification A 5.1.
The chemical requirements of A 5.1 address the reporting of Manganese, Silicon, Nickel, Chromium, Molybdenum, and Vanadium.
The certification aadresses Carbon, Nan-ganese, Sulfur, Phosphorus, and Silicon, thus excluding four required elements.
45
ORGANIZATION:
COMSIP, INC.
CUSTOMLINE DIVISION LINDEN, NEW JERSEY REPORT INSPECIl0N p
NO.;
99900771/82-01 RESULTS:
PAGE 4 of 5 6.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and QA Manual Procedure 6.4, paragraph 5.5, the following examples were observed of different drawing revision numbers being used in the manufacturing process to those listed on the Project Drawing Control Board, and the affected assemblies were not identified as being nonconforming:
Drawings listed on Drawings being used Project Drawing Control Board in Manufacturing 5875-17 Sheet 6, Rev. 5 5875-17 Sheet 6, Rev. 4 5875-17 Sheet 7, Rev. 6 5875-17 Sheet 7, Rev. 5 5875-17 Sheet 17, Rev. 6 5875-17 Sheet 17, Rev. 5 5875-17 Sheet 19, Rev. 11 5875-17 Sheet 19, Rev. 10 5875-17 Sheet 20, Rev. 9 5875-17 Sheet 20, Rev. 8 5875-17 Sheet 21, Rev. 8 5875-17 Sheet 21, Rev. 7 It should be noted that the above drawings had been issued to the shop supervisor by the drawing distribution control clerk, however, the shop supervisor had not issued the latest revisions to the shop personnel.
C.
UNRESOLVED ITEMS:
None D.
OTHER FINDINGS OR COMMENTS:
1.
NRC Region III Inspection Request - This area of the inspection was conducted by review of records pertaining to the main control panels provided to the Callaway Plant, Unit 1, and review and observation of current welding activities related to the fabrication of the main control panels being supplied to Palo Verde Nuclear Generating Station, Unit 3.
The welding procedure specification and its quali-fications, welder qualifications, welding material control, and welding inspection were reviewed, and observations made of inprocess welding activities.
As a result, nonconformance B.1 through B.5 were identified.
Comment:
In regard to nonconformance B.4, visual inspection for other types of defects in v. elds was impaired by the presence of a paint film of up to 6 mils thickness.
1 46
ORGANIZATION:
COMSIP, INC.
CUSTOMLINE DIVISION LINDEN, NEW JERSEY RLPORI IN5PtLi10H NO.:
99900771/82-01 RESULTS:
PAGE 5 of 5 2.
Drawing Control: - This area of the inspection was conducted by review of the Project Drawing Control Board and Drawing Status Log, which identify the latest, approved project drawings, and comparing against the drawings actually in use in the manufacturing area.
Further, the implementation of controls for initiating drawing revisions by Design Control Change Notices was reviewed.
As a result of this inspection, nonconformance B.6 was identified.
47
ORGANIZATION:
CORNER AND LADA COMPANY, INCORPORATED CRANSTON, RHODE ISLAND RLPUni INSPECIION INSPECTION NO.
99900349/82-01 DATE(S) 4/5-8/82 ON-SITE HOURS: 24 CORRESPONDENCE ADDRESS:
Corner and Lada Company, Incorporated ATIN:
Mr. W. T. Allen, III Manager, Quality Assurance 1341 Elmwood Avenue Cranston, Rhode Island 02910 ORGANIZATIONAL CONTACT:
Mr. W.
T. Allen III TELEPHONE NUMBER:
(401)461-1300 PRINCIPAL PRODUCT: Nuclear Component Supports NUCLEAR INDUSTRY ACTIVITY: Commercial nuclear production of Corner and Lada Co.,
Inc., Cranston, Rhode Island represents 40% of total company production.
Five nuclear contracts are presently in-house for current and future production.
Cf mD
~~ ?
C-1 r2 ASSIGNED INSPECTOR:
4
/+
J. W. Sutton, Reactive & Components Program Section Date (R& CPS)
OTHER INSPECTOR (S):
APPROVED BY:
$..W
. e. 4 2.
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
Status of previous identified findings, nonconformance/ corrective action, rev ew of vendor activities.
4 PLANT SITE APPLICABILITY:
Not identified.
49
ORGANIZATION:
CORNER AND LADA COMPANY, INCORPORATED CRANSTON, RHODE ISLAND REPORT INSPECTION NO.-
99900349/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Violation (80-01) - Failure to comply with 10 CFR 21 relative to evaluation of possible deviations and notification to purchasers of parts with possible deviations.
The inspector reviewed the corrective action taken by Corner and Lada (C&L) to prevent recurrence.
The inspector reviewed for content the revised standard specification procedure ST-CFR-21, documentation of the training session held on February 17, 1981, and the corrective action request issued on January 30, 1981. In addition, the inspector reviewed 10 letters sent by C&L to affected customers and the load limit evaluation performed by C&L to determine if a safety hazard existed.
As a result of this review, the inspector determined that adequate corrective action and preventive measures had been taken by C&L to prevent recurrence of this reported violation.
2.
(Closed) Deviation A (80-01) - C&L not reviewing and approving drawings as required by procedure ST-150.
The inspector reviewed the revised corrective action request, two design drawings, 9 drawing change requests and the cross log revision documentation, to determine adherence to C&L's corrective action commitments.
The inspector determined that the corrective action taken will prevent recurrence of this problem.
50
ORGANIZATION:
CORNER AND LADA COMPANY, INCORPORATED CRANSTON, RHODE ISLAND l
l REPORT INSPECIlON I
NO.
99900349/82-01 RESULTS:
PAGE 3 of 3 1
3.
(Closed) Deviation B (80-01) - Reamed holes in sway strut paddles do not meet drawing requirements.
The inspector reviewed the corrective action taken by C&L to prevent recurrence of this problem.
The measurement instruments purchased were examined and found to be acceptable.
The inspector reviewed revisions to drawings indicating dimensional changes to the reamed hole in the rigid sway strut paddle.
As a result of the review of C&L corrective action, the inspector determined that the corrective action taken by C&L will prevent recurrence of this deviation.
E.
OTHER FINDINGS:
1.
Nonconformance/ Corrective Action - The inspector reviewed Section 10 of the QA Manual, 18 nonconformance material reports, 20 corrective action reports and associated documentation, to determine compliance to QA/QC requirements.
Segregation areas designated for storage of nonconforming items were inspected.
Relative to the documents examined, the inspector determined that activities relating to the control and disposition of nonconforming items complied with C&L QA/QC program requirements.
51
ORGANIZATION:
DRAVO CORPORATION PIPE FABRICATION DIVISION MARIETTA, OHIO REPORi INSPECIION INSPECIION NO.
99900017/82-01 DATE(5) 3/29-4/2/82 UN-SITE HOURS: 78 CLRRESPONDENCE ADDRESS:
Dravo Corporation Pipe Fabrication Division ATTN:
K. Anderson, General Manager 1115 Gilman Street Marietta, Ohio 45750 ORGANIZATIONAL CONTACT:
W. A. Molvie, Quality Assurance Manager TELEPHONE NUMBER:
(614) 373-7541 PRINCIPAL PRODUCT: Nuclear Piping Subassemblies NUCLEAR INDUSTRY ACTIVITY: Approximately 10% of production devoted to nuclear products.
2 ASSIGNED INSPECTOR: y A'l Muh
/If/b-l I
H. W.
Rolle'rds, Reactive and Components Program Date' Section (R& CPS)
OTHER INSPECTOR (5): J. Conway, R& CPS a%
S / w /# 2_
APPROVED BY:
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
SCOPE:
This inspection was made as the result of:
(1) a 10 CFR Part 50.55(e) report by TVA relative to undocumented weld buildups on ASME Section III Class 1, 2, and 3 pipe ends supplied to Watts Bar Nuclear Plant, Units 1 and 2; and (2) the identification at Callaway, Units 1 and 2 of inadequate surface preparation of safety-related piping welds for performance of liquid penetrant inspection.
PLANT SITE APPLICABILITY:
Undoce..ented weld buildups 390, and 50-391.
Inadequate weld surface preparation for liquid penetrant inspection 483, and 50-486.
53 l
ORGANIZATION:
DRAVO CORPORATION PIPE FABRICATION DIVISION MARIETTA, OHIO REPORT INSPECTION NO.
99900017/82-01 RESULTS:
PAGE 2 of 4 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 9 of the QA Manual, certain inspection and verification operations on the weld joint data forms, applicable to the attachment of nameplates to piping subassemblies, had been signed-off by the weld foreman, as completed, although the name-plates were now attached and were observed in the record package.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 15.4.2 and 15.4.3 of Dravo's Nuclear QA Manual, corrective action implementation date and verification followup were not docu-mented with respect to an internal audit finding pertaining to use of unapproved welding wire.
C.
UNRESOLVED ITEMS:
None D.
OTHER FINDINGS OR COMMENTS:
1.
Inadequate Surface Preparation For Liquid Penetrant Inspection of Piping Welds Fabricated by Dravo - As the result of a United Engineers &
& Constructors (UE&C) Surveillance Representative observing a Dravo inspector performing an improper magnetic particle examination on an ASME Section III piping subassembly for the Seabrook Nuclear Station, Dravo notified Callaway, Wolf Creek Generating Station, and Seabrook Nuclear Station of a potential deficiency relative to incomplete exami-nation of certain welds.
In the process of reexamination of welds on Dravo piping subassemblies by contractor personnel at Callaway, it was reported that liquid penetrant examination could not be performed because of inadequate surface preparation of welds.
As a result of the above deficiencies Dravo has taken the following corrective action and steps to preclude recurrence:
54 1
ORGANIZATION:
DRAVO CORPORATION PIPE FABRICATION DIVISION MARIETTA, OHIO
^
REPORT IN5PLCI10N NO.
99900017/82-01 RESULTS:
PAGE 3 of 4 a.
The inspector responsible for the inadequate inspection resigned from the company and all assemblies in the Dravo shop that had inspections performed by this individual were reinspected, b.
Training sassions have been given to all inspection personnel relative to ASME Code and Dravo procedure requiements.
c.
Standards have been developed to demonstrate inspection surface requirements.
d.
The Chief Inspector (Level III Examiner) monitors the work of each Dravo inspector on a weekly basis, and records are main-tained to attest to this surveillance.
The inspection revealed that other nuclear facilities could have similar deficiencies but were not included in Dravo's contract review, in that, the review was made on only work performed during the last hal f of 1980.
2.
Nuclear Regulatory Commission, Region II Referral of Undocumented Weld Buildup on Safety-Related Piping Fabricated by Dravo -
a.
In February 1980, TVA notified Dravo that they had noted indica-tions on butt welds while performing UT (baseline inspection).
Both TVA and Dravo agreed that the indications were the result of weld buildups made in the shop.
These weld buildups had not been documented in accordance with Code requirements.
Dravo was able to reconstruct from fabrication records the documentation required for the identified suspect welds with the exception of 14 which had 0.D. buildups.
In September 1980, Dravo sent TVA a nonconformance report documenting the 14 sketches for the Watts Bar contract and recommending that the pipe assemblies containing the undocumented weld buildups be "used as is."
Dravo's corrective action to preclude undocumented weld buildups in the future included revised procedures to incorporate better control of weld material and a QA review of weld joint data cards.
In addition ECN-1 " Instructions for Build-Up" was written to include the required nondestructive examinations and documentation for weld buildups on all ASME Class 1, 2, and 3 work.
ORGANIZATION:
DRAVO CORPORATION PIPE FABRICATION DIVISION MARIETTA, OHIO REPORI INSPEGi10N NO..
99900017/82-01 RESULTS:
PAGE 4 of 4 b.
On May 28, 1980, the engineering organization generated a devia-tion request relating to the weld " buildup" problem and recommendea corrective action to establish an Engineering Change Notice (ECN) system and incorporate it into the QA Manual.
There was no evidence that the QA organization acknowledged the deviation request nor followed through on implementation of the recommended corrective action.
This item will be further reviewed during a future inspection.
c.
An NRC inspection of weld buildup documentation was performed by comparison of pipe cutting sheets (which make provisions for indicating whether a buildup is required) with subassembly fabri-cation records.
Subassemblies from Seabrook, Units 1 and 2, Wolf Creek, and Hope Creek, Units 1 and 2, were included in the inspection sample.
Fabrication records showed absence of docu-mentation of buildup being performed in only one instance, from a total of six subassemblies indicated by cutting sheets to require buildup; i.e., Hope Creek, Sketch E3035-117, Class 2.
1 i
Examination of the specific requirements applicable to the TVA Watts Bar contract showed that the failure to document certain weld buildups at pipe ends was related to QA program implementa-tion rather than failure to define requirements.
Review of General Fabrication Procedure SP-1 for this contract indicated buildups were required to be reported to the Engineering Department, in order that Shop Sketches would be appropriately revised.
f 56
ORGANIZATION:
ENERGY INCORPORATED IDAHO FALLS, IDAHO RLPORI INSPECTION INSPECTION NO.
99900514/82-01 DATE(S) 4/6-8/82 ON-SITE HOURS: 24 CORRESPONDENCE ADDRESS:
Energy Incorporated ATTN:
Mr. W. V. Botts President and Chairman P. O. Box 736 Idaho Falls, Idaho 83401 ORGANIZATIONAL CONTACT:
T. E. Didesch, QA Manager TELEPHONE NUMBER:
(208) 529-1000 PRINCIPAL PRODUCT: Engineering Consultants NUCLEAR INDUSTRY ACTIVITY: Approximately 43% to 64% of the Energy Incorporated staff of 280 personnel are assigned to safety-related projects for the nuclear industry.
\\
r ASSIGNED INSPECTOR:
~
,1
//-
- / T. ' / /,
R. H. Brickley, Reactor Systems Section (RSS)
Date OTHER INSPECTOR (S): J. L. Carter, Reactor Systems Branch (RSB)/NRR J. Guttmann, RSB/NRR T. Y. C. Wei, Argonne National Laboratory APPROVED BY:
\\
[' ' A-h
_ /'.' i /
j C. J. Hale, Chief, RSS Date '
INSPECTION BASES AND SCOPE:
A.
BASES:
SCOPE:
Initial management :neeting and a special inspection in response to a request from NRR, regarding validation of the computer code RETRAN.
PLANT SITE APPLICABILITY:
Not identified.
S7
ORGANIZATION:
ENERGY INCORPORATED IDAHO FALLS, ICAHO RLPORI INSPECilON NO..
99900514/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
Contrary to the requirements of Section 21.6 of 10 CFR Part 21, Energy incorporated (EI) did not have posted:
(1) current copies of Section 206 of the Energy Reorganization Act of 1974, (2) the name of the individual to whom reports may be made, arid (3) a state-ment as to where the regulations / procedures may be examined.
B.
NONCONFORMANCES:
Contrary to the requirements of Criterion V of 10 CFR Part 50, Appendix B and Section III of EI Standard Operating Procedure IX-I (Compliance with 10 CFR Part 21), El purchase orders No. 14896, for seismic testing of safety-related equipment, and No. 14920, to fabricate isolation signal logic decoder boards classified as safety-related, did not specify that the provisions of 10 CFR Part 21 apply.
C.
UNRESOLVED ITEMS:
None D.
OTHER FINGINGS OR COMMENTS:
1.
Initial Management Meeting - A meeting was conducted with the management of Energy Incorporated to acquaint them with the Licensee Contractor Vendor Inspection Program (LCVIP), the NRC organization relative to the Vendor Programs Branch, and the types of documents generated and processed in implementing the program.
2.
RETRAN Computer Code - This special inspection is part of a series of inspections being conducted at the request of NRR regarding the development, verification, and use of the RETRAN computer code by the nuclear industry.
The basic objectives of these inspections are to assure that the RETRAN computer code has been developed, maintained, and modified in accordance with the design control measures contained in ANSI N45.2.11 - 1974; and to verify that the organizations' computer code activities address the requirements of 10 CFR Part 21, 10 CFR Part 50.55(e) and 50.59, as applicable.
j 58
ORGANIZATION:
ENERGY INCORPORATED IDAHO FALLS, IDAHO RtPORI INSPtGilOh NO..
99900514/82-01 RESULTS:
PAGE 3 of 3 A review of the RETRAN files disclosed that the code had not been developed, maintained, and modified in accordance with the EI Quality Assurance Manual and its implementing procedures.
In addition, a review of the contract agreement between the Electric Power Research Institute and EI disclosed that no QA program requirements had been imposed on these activities.
El personnel stated that RETRAN was developed and maintained in accordance with the EI QA program; however, there was no formal documentation of their activities, as would be required by the QA program.
The current EI QA program and its implementing procedures were inspected as they relate to computer codes activities.
Eleven sections of the Quality Assurance Manual, five sections of the Engineering Procedures Manual, seven sections of the Software Procedures Manual, and one Standard Operating Procedure (50P) were reviewed to determine their conformance with the applicable requirements of ANSI N45.2.ll-1974 and NRC reporting requirements.
A verification of the degree of implementation of 10 CFR Part 21 was obtained by examination of the 50P to implement 10 CFR Part 21, one official bulletin board, and three purchase orders.
The review of the EI QA program and implementing procedures disclosed that minor revisions are needed to the EI Quality Assurance Manual and its implementing procedures for clarification purposes; e.g.,
restric-tions on use of supervisor as independent reviewer; defining terms like periodic and significant; and use of correct checklists for input and independent reviews.
The specific items needing revision were discussed with cognizant El management.
The incorporation of these items into the QA program and their implementation will be reviewed during a future inspection.
The inspection of the implementation of 10 CFR Part 21 requirements resulted in the violation and nonconformance identified in A and B above.
In regard to the violation iGentified in A above, it should be noted that:
(1) a portion of the complete text of Section 206 of the Energy Reorganization Act of 1974 had been omitted; and (2) the individual to whom reports may be made differed from the one identified in the 50P.
59
ORGANIZATION:
EXIDE CORPORATION RICHMOND, KENTUCKY REPORT INSPECTION INSPECTION NO.
99900359/82-01 DATE(S) 3/16-19/82 ON-SITE HOURS: 27 CORRESPONDENCE ADDRESS:
Exide Corporation j
ATTN:
Mr. Charles Reichart Contracts Administrator 101 Gibraltor Rd.
Horsham, PA 19044 ORGANIZATIONAL CONTACT:
Mr. Earl Price, Plant QA Manager l
TELEPHONE NUMBER:
(606) 624-1000 PRINCIPAL PRODUCT: Lead Acid Batteries NUCLEAR INDUSTRY ACTIVITY: The Richmond, Kentucky, plant nuclear output, based on sales is 8%.
,/
/ D c.-
A~ o 2 ASSIGNED INSPECTOR:
,.4 R. E. Oller, Reactive & Components Program Date Section (R& CPS)
OTHER INSPECTOR (S):
APPROVED BY:
-f, / )
'i 6 --
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
1.
BASES:
2.
SCOPE:
Status of previous inspection findings; manufacturing process control; change control, and nonconformances and corrective action.
PLANT SITE APPLICABILITY:
Perry Nuclear Station, Units 1 and 2, Dockets 50-440 and 50-441 61
ORGANIZATION:
EXIDE CORPORATION RICHMOND, KENTUCKY REPORI INSPtCl10N NO.
99900359/82-01 RESULTS:
PAGE 2 of 6 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Deviation A (79-01):
Failure to include the purchase order number in the identification of 60 type 2 GN-15 batteries, as required by paragraph 8.3.7 of the QA Manual, Revision 8.
The inspector verified that in accordance with the vendor's response letter, dated June 19, 1979, review by Exide had established that the stationary battery post design could not accommodate the application of the F.0. number.
To resolve this problem, Exide has revised paragraph 8.3.7 of the QAM to delete l
this requirement.
The current practice is to identify the
}
batteries by stamping the posts with codes traceable to test data reports which identify the type, catalog designation, location, shipping date and P.O. number.
2.
(Closed) Deviation B (79-01):
Failure to provide MRB hold areas for segregation of nonconforming manufactured parts, subassemblies, and assemblies, as required by paragraph 15.9 of the QA Manual, Revision 8.
The inspector verified by observations in the shop, that in accordance with the vendor's response letter, dated June 19, 1979, four MRB hold areas are now identified in the shop with signs and boundary paint anarks.
Items placed in these areas are red tagged and monitored on a daily basis by the assigned QC area controller.
62
ORGANIZATION:
EXIDE CORPORATION RICHMOND, KENTUCKY REPORI INSPtLiiUN NO.
99900359/82-01 RESULTS:
PAGE 3 of 6 3.
(Closed) Deviation C (79-01):
Failure to define and document the effective date of corrective action for Corrective Action Requests Nos. R-006 and R-013, as required by paragraph 16.4.2 of the QA Manual, Revision 8.
The inspector verified that in accordance with the vendor's response letter, dated June 19, 1979, the two above CARS and current CARS had the corrective actions defined, signed, and dated by the plant QC Manager.
Discussion, however, established that the requirements of paragraph 16.4.2 of the QAM were not clear, and the Exide QA Engineer indicated this paragraph will be revised to more clearly define this requirement.
4.
(Closed) Deviation D (79-01):
Failure to perform and document yearly audits in the areas of M/P cell assembly, stationary cell assembly, and plate finishing prior to March 30, 1979, and of the areas of casting / pasting / filling after August 25, 1977, as required by paragraph 18.3.1 of the QA Manual, Revision 8.
The inspector verified that in accordance with the vendor's response letters dated June 19 and November 1, 1979, the entire Richmond plant QA program was audited on June 20, 1978, by the Chief Engineer, and on April 13, 1979, by the Manager, Design Engineering, and more recently by a management audit team.
The inspector also determined that while procedure QCP-57.0, which controlled the audits by QA management, listed nine production activities which would be audited on an annual basis, the assigned checklist used by the auditors, did not specifically address the nine areas.
The checklist, instead, addressed the entire QA program, and it could not be specifically related to the nine production activities.
The Exide QA Engineer indicated he would revise the checklist to clarify this condition.
5.
(Closed) Deviation E (79-01):
Failure to maintain the required melt time of 15 seconds and a mating time of 20 seconds for stationary cells, as required by paragraph 3.8.7.a of the Temporary Manufacturing Requirement No.14.1(R), dated May 16, 1978, and failure to air test the stationary cells at both 1.5 and 2.0 psi as required by Special Manufacturing Instruction No. 6, dated July 26, 1978.
63
ORGANIZATION:
EXIDE CORPORATION RICHMOND, KENTUCKY REPORT IN5FtL61UN l
NO.
99900359/82-01 RESULTS:
PAGE 4 of 6 The inspector verified that in accordance with the vendor's response letters dated June 19 and November 1, 1979, the TMI No. 14.l(R),
dated May 16, 1978, was in error and it was revised on May 9, 1979, to correct the melt and mating time to minimum values of 15 and 20 seconds, respectively.
With regard to the air test, the SMI No. 6 was cancelled and superseded by the revised TMI which specified the air test shall be performed only at 1.5 psi for 15 seconds minimum.
This TMI was subsequently superseded by Manufacturing Requirement No. 102, Section 20.6, issued August 24, 1981.
These activities are currently being monitored on a weekly basis by the QC area controllers and recorded on checklists.
6.
(Closed) Deviation F (79-01):
Failure to indicate specific hold points in documents controlling manufacturing, inspection, and test activities, to accomodate customer imposition of mandatory inspection hold points, as required by paragraph 10.12 of the QA Manual, Revi-sion 8.
The inspector verified that in accordance with the vendor's response letters dated June 19 and November 1, 1979, the procedure controlling establishment of hold points is No. QCP 60.0 and a " Customer Contract-Quality Assurance Checklist." This QCP and checklist were revised on November 1, 1979, to reflect organizational changes.
They were i
subsequently superseded by procedure QAP-60.0 on September 22, 1981.
Review by the inspector of QAP-60.0 and a current Customer Contract QA/C Checklist for Exide Order No. 77706/77709 for Perry Nuclear Station established that this checklist provided for Customer Witness and Hold Points.
I 7.
(0 pen) Unresolved Item E.3.b.1 (79-01):
Paragraph 7.11 of the QA Manual, Revision 8, and paragraph 4.2 of the QC Procedure No. 30.1, dated June 13, 1977, require that the checker at receiving initial each invoice upon receipt of purchased material.
In practice, the checker initials the Bill of Lading and initiates a Receiving Report.
The inspector verified that the statements in the QA manual and QA procedure were in error since the invoices do not go to the receiving checker, but do go directly to the accounting department.
The Exide QA Engineer indicated that the QA manual and QC procedure will be revised to reflect the actual practice.
64
ORGANIZATION:
EXIDE CORPORATION RICHMOND, KENTUCKY REPORI Ih5PhLl10N NO.-
99900359/82-01 RESULTS:
PAGE 5 of 6 8.
(Closed) Unresolved Item E.3.b.2 (79-01):
Due to the area identifi-cation on the audit checklist being different from the audit area identification in the QC procedure, the inspector was unable to determine that areas were audited as required.
The inspector found during the current inspection that this item is a duplicate of deviation D and is closed.
E.
OTHER FINDINGS OR COMMENTS:
1.
Manufacturing Process Control:
The inspector reviewed seven sections of the Exide QA manual which were applicable to the GN (nuclear) type stationary battery parts manufacture, assembly, tests, inspections, QC monitoring, and shipping.
Observations were made of battery grids and small parts casting, grid pasting and finishinp, caemical sampling, battery assembly, final inspection, and the calibration status of production measuring tools and process control indicating and recording instruments.
Review, also, included the following QC monitoring records, procedures, and specifications maintained in various manufacturing areas by the four QC Area Controllers:
10 QC procedures; 7 manufacturing require-ments procedures; 3 quality characteristic lists specifications; 1 process specification; 8 weekly castings QC checklists; 10 plate pasting QC checklists; 10 weekly stationary inspection and process verification QC checklists; 4 weekly stationary product inspection reports; 10 weekly battery finishing and shipping QC checklists; 10 weekly small parts process QC checklists; and 6 (twice weekly) charging process QC checklists.
Within this area, no nonconformances to NRC or QA program requirements were identified, and no unresolved items were identified.
2.
Change Control:
The inspector reviewed the QA Manual Section 3,
" Design Documentation and Change Control," and Section 6, " Quality Documents, Preparation and Issuance Of," which were applicable to changes in engineering generated manufacturing documents, and QA and QC documents.
A review was also made of:
an administrative document control procedure QAP-55.0; two records of Engineering Change Notices requesting changes to manufacturing requirement procedures; one record of an Engineering Specification Notice; one Design Notice covering a change in a design drawing; and one record of a Purchase Specification Notice.
65
ORGANIZATION:
EXIDE CORPORATION RICHMOND, KENTUCKY REPORT INSPECI10N
(
NO..
99900339/82-01 RESULTS:
PAGE 6 of 6 Review and discussion also established that only the QC procedures and their changes, were totally generated and approved at the Richmond plant level.
While changes to engineering, manufacturing, and QA documents may be requested at plant level using ECRs, the final 6pprovals and changes are controlled at the Yardley and Horsham, Pennsylvania, engineering, quality assurance, and corporate levels.
Within this area, no nonconformances to NRC and QA program require-ments were identified and no unresolved items were identified.
3.
Nonconformances and Corrective Action:
The inspector reviewed two sections of the Exide QA manual which were applicable to control and disposition of deficiencies found in materials at receiving inspection, parts during manufacture and assembly, and batteries during final inspection prior to shipping.
This control of noncon-forming items involved those items which could not be dispositioned during normal processing under the red tag system.
To verify that nonconforming items were reviewed and accepted, rejr ted, repaired, or reworked, and that conditions adverse to qur iity, including the cause were corrected, a review was made of:
2 administrative QA procedures; 10 Material Review Reports common to the GN (nuclear) type batteries; 4 Vendor Defective Material Reports; and 2 Corrective Action Requests.
Observations were made of four hold areas in the shop, designated for items on hold for the Material Review Board disposition.
Within this area, no nonconformances to NRC and QA program require-ments were identified, and no unresolved items were identified.
66
ORGANIZATION:
THE FOXBORO COMPANY HIGHLAND PLANT EAST BRIDGEWATER, MASSACHUSETTS l0N-SITEHOURS:
INSPECIl0N REPORI INSPECIl0N 24 NO.-
99900225/82-01 DATE(5) 4/6-8/82 CORRESPONDENCE ADDRESS:
The Foxboro Company ATTN:
Mr. C. A. McKay, Executive Vice President Technology Division 38 Neponset Ave.
Foxboro, MA 02035 ORGANIZATIONAL CONTACT:
Mr. F. Leathers, Corporate QA Manager, Field Operations TELEPHONE NUMBER:
(617) 543-8750 PRINCIPAL PRODUCT: Electronic Process Control Instrumentation and Allied Equipment.
NUCLEAR INDUSTRY ACTIVITY: The Highland Plant Nuclear production output is 2.5% based on sales.
7f [
b4 5 -2 7-J2 ASSIGNED INSPECTOR:
R. E. Oller, Reactive & Component Program Section Date (R& CPS)
OTHER INSPECTOR (S):
,M,h ' 8
r/w/f'2__
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
General Electric Topical Report NED0-11209-04.
B.
SCOPE:
Procurement controls, cladding fabrication, fuel rod fabrication, and status of previous inspection findings.
PLANT SITE APPLICABILITY:
Not identified.
73
ORGANIZATION:
GENERAL ELECTRIC COMPANY WILMINGTON MANUFACTURING DEPARMENT WILMINGTON, NORTH CAROLINA REPORT INSPECIl0N
(
NO.
99900003/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
Contrary to Section 5 of the NRC approved Topical Report NEDO-11209-04 and paragraph 5.5 of Product / Process Quality Plan FC0 50.2, there was no Quality Notice or other data to document the temperature profile verification in 1981 for four of five autoclaves.
There was no Quality Notice but there was data to document that temperature profiles were verified for three of the autoclaves in 1980.
C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (81-01):
The applicable project code effectivity date was not included in a material requisition and its subsequent purchase order.
A review of current purchase orders and material requisitions uncovered no further problem in this area.
GE has revised its Code Manual to describe the order entry documents as the key source of code effectivity information.
2.
(Closed) Nonconformance (81-02):
Fluorine, chlorine, carbon, and nitro-i gen samples, used by the computer to calculate total impurities and I
boron equivalence, were not always drawn from the same material as the metal impurity sample for Gadolinium fuel.
A review of current release lot found the samples drawn from the same tray of pellets.
The Quality Control Instruction and the computer program have been revised to require samples to be drawn from the same tray.
E.
OTHER FINDINGS OR COMMENTS:
1.
Follow up on Unresolved Item - Verification of whether the material on Furnace Sheet 1573, which was suspected to have high hydrogen, had been placed on hold in the Manufacturing Information and Control System (MICS).
GE found that it could not recall the historical information from its computerized MICS system.
Computer inputs have now been coded as historical information.
The recall of historical information on the acceptance ststus of a current lot was verified.
This unresolved item is now closed.
74
ORGANIZATION:
GENERAL ELECTRIC COMPANY WILMINGTON MANUFACTURING DEPARTMENT WILMINGTON, NORTH CAROLINA REPORI Hoettlion NO.
99900003/82-01 RESULTS:
PAGE 3 of ?
2.
Procurement Controls - The procurement controls used on Kearsarge, an investment casting subvendor, were reviewed.
Additional controls and inspection had been placed on Kearsarge before it discontinued business.
Several purchase orders for various parts, such as velocity limiters and fuel supports, and their associated drawings and specifications, were reviewed.
In addition, the surveillance reports and receipt inspection activities were reviewed.
The nonconformances reported by Kearsarge to GE were verified to have been processed by GE in accordance with established procedures.
3.
Cladding Fabrication - The specifications, drawings, and quality plans related to cladding fabrication were reviewed.
The release of recent lots of cladding was inspected.
The results of chemical and physical testing were verified to be in conformance with the
- 6 specications and quality plans requirements.
The processing of cladding, in particular, autoclaving and annealing were inspected.
The qualification of the processes and personnel was verified and the conformance of the processes to specifications and quality plans was verified.
In this area, one nonconformance was identified.
Before the end of the inspection, GE began the corrective action of performing an autoclave thermocouple run.
4.
Fuel Rod Fabrication - The fuel design docume:'c specifications, drawings, and etc., were reviewed.
The personnel and station qualifi-cations were verified.
The welding of plugs to fuel clad was witnessed.
In such, the conformance to specification and procedure requirements was verified.
75
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA RLPORI INSPECTION INSPECTION NO.
99900403/82-01 DATE(5) 2/22-26/82 ON-SITE HOURS: 58 CORRESPONDENCE ADDRESS:
General Electric Company Nuclear Energy Business Operations ATTN:
Mr. W. H. Bruggeman Vice President and General Manager 175 Curtner Avenue San Jose, California 95125 ORGANIZATIONAL CONTACT:
Mr. A. Breed, Manager, Quality Assurance TELEPHONE NUMBER:
(408) 925-2726 PRINCIPAL PRODUCT: Nuclear Steam System Supplier NUCLEAR INDUSTRY ACTIVITY: General Electric Company, Nuclear Energy Business Operations (NEBO), has a work force of approximately 7650 peop e with approxi-mately 98% of that work force devoted to domestic nuclear activity.
NEB 0 currently has 26 reactor units under construction and 4 reactor units under i
contract.
NEB 0 also has approximately 115 service contracts with various l
clients.
o /Av 5/J6/TA_
ASSIGNED INSPECTOR:
D. D. @ derlain7 ke/ftor Systems Section (RSS)
D a te'
(
OTHER INSPECTOR (S): W. E. Foster, Reactive & Components Program Section APPROVED BY:
i Q
M.%/ A
~
C. J @ le, thief)SS Date #
INSPECTION BASES AND SCOPE:
A.
BASES:
General Electric Topical Report No. NE00-11209-04A and 10 CFR Part 50, Appendix B.
B.
SCOPE:
See next page.
PLANT SITE APPLICABILITY:
Docket Numbers 50-416/417, 50-466/467, 50-556/557, 50-461/462, 50-367, 50-354/355, 50-410, 50-440/441, 50-458/459, 50-522/523, 50-373/374, 50-373/374, 50-466/467, 50-553/554, and 50-518/519/520/521.
77 l
1
l ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR CNERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION NO.
99900403/82-01 RESULTS:
PAGE 2 of 10 B.
SCOPE:
Status of previous inspection findings and follow up on the following requests:
1.
Allegations that General Electric (GE) San Jose had sent technical publications to licensees containing known errors.
2.
Region III request regarding potential document control problems at GE (LaSalle project).
3.
10 CFR Part 50.55(e) report (River Bend project) stating that the heat loads provided by GE to Stone & Webster (S&W) for sizing the heating and ventilating (HVAC) system in the high pressure coolant injection (HPCI) system diesel generator room were about one-third of the expected value.
4.
10 CFR Part 50.55(e) report (TVA) stating that GE was not performing 50.55(e) evaluation of potential reportable conditions as required.
Additionally, this inspection was conducted as a result of Mississippi Power and Light Company's issuance of 10 CFR Part 50.55(e) reports pertaining to (1) inadequate circuit separation in Power Generation Control Complexes (PGCC), (2) incorrect assembly of PGCC cable connectors, (3) termination of size 8 AWG and smaller wire in junction boxes of PGCC termination cabinets, (4) inability to adjust transmitters and trip units for the reactor vessel water level scrams, (5) improper contact arrangement of Series 20, Type PR-20 Electro Switch, and (6) incompatibility between design documents and termina-tion cabinets.
Items (5) and (6) were partially evaluated during a previous inspection.
All items refer to hardware supplied to the Grand Gulf Nuclear Generating Station.
78
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA i
REPORT INSPELI10N NO.
99900403/82-01 RESULTS:
PAGE 3 of 10 A.
VIOLATIONS:
None l
B.
NONCONFORMANCES:
1.
Contrary to Section 5 of Topical Report No. NED0-11209-04A, Revision 3, dated August 1981, and paragraph 5.1 of Fire Test Specification No. A00-626 (Design Record File No. A00-794-6),
the final test report for the fire test had not been signed by the performer.
Prior to the close of this inspection, the inspector was provided a copy of a letter that gave an interpretation of the headings identified on the test report cover sheet.
Additionally, the test report cover sheet was modified by adding a heading entitled "Per-formed By" This heading was completed by the typed name and signature of the performer and the date.
Since corrective and preventive actions were taken before the end of the inspection, no further written response is required.
2.
Contrary to Section 5 of Topical Report No. NED0-11209-04A, Revision 3, dated August 1981, and paragraph 5.3.3 of Manufacturing Standard Practice (MSP) No. 10.003, Revision 3, dated February 26, 1981, Section 14 of the Power Generation Control Complex Cable Assembly Instruction Manual had not been released to the station where the specified task (connector preassembly) was being I
accomplished.
The above paragraph of MSP 10.003 indicated that individual sections may be released to the floor for use in cable manufacturing.
It was apparent that responsible individuals elected to release sections to the floor, as evidenced by appropriate sections at work stations.
l 79
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION i
NO.
99900403/62-01 RESULTS:
PAGE 4 of 10 Prior to the close of this inspection, the inspector was provided a copy of a letter and its attached Corrective Action Request, a sample Surveillance Program Check Sheet, and an " Advanced Copy" of MSP No. 10.003, Revision 4, dated February 25, 1982.
Since corrective and preventive actions were taken before the end of this inspection, no further written response is required.
3.
Contrary to Section 5 of Topical Report No. NE00-11209-04A, Revision 3, dated August 1981, and paragraph IV.B of Nuclear Energy Business Group Procedure No. 70-42, Revision 4, dated March 27, 1981, requested additional information, or a schedule for such information, had not been provided within 5 working days after receipt of the request.
The request was made to Quality Assurance in a memorandum, dated May 27, 1981, with a required response date of June 5, 1981.
The memorandum trans-mitting the requested information was dated June 26, 1981.
The date that Quality Assurance received the memorandum of May 27, 1981, is not verifiable because it was not date stamped upon receipt.
However, it was received prior to June 5, 1981, as evidenced by its (subject matter /PRC No.) inclusion in a Quality Assurance status report dated June 5, 1981.
4.
Contrary to the requirements of 10 CFR Part 50, Appendix B and procedure E0P 30-10.00 Section 1.6, no Engineering Review Memorandum (ERM) existed to document the required reviews of GEK-75662A -
Clinton 1 - Operation and Maintenance Instructions - High Pressure Core Spray System.
I C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
(Closed) Nonconformance (81-03):
Material Requests (MRs) for safety-related components were not being accomplished and controlled by established procedures and standing instructions.
Quality Control Standing Instruction (QCSI) 7.2.18 and Engineering Operating Procedure (EOP) 42-5.00 were revised to strengthen and clarify program requirements for tracking, review, and filing of MRs and Purchase Orders (P.O.).
Also, an internal management memo was directed to all QC Engineers concerning the need for strict compliance with QCSI 7.2.18 regarding MR/P0 control.
80
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORI IN5PELfl0N NO.
99900403/82-01 RESULTS:
PAGE 5 of 10 E.
OTHER FINDINGS OR COMMENTS:
1.
Technical Publications to Licensees Containing Known Errors - This area of inspection resulted from a Region V request for follow up at GE on allegations regarding technical publications, containing known errors, being sent to licensees.
The concerns related to Operation and Main-tenance (0&M) Instruction Manuals for GE supplied systems.
Applicable procedures were examined to determine program requirements relative to O&M manuals.
The following documents were examined in order to verify implementation of the program requirements:
a typical list of standard NSSS O&M instruction manuals; nine issued O&M manuals; and four Engineering Review Memoranda which document the required engineering review of the 0&M manuals.
The review ranged from a spot check (for such things as control rod drive voltages) of the issued O&M manuals against the draft copy that was reviewed by engineering to a word for word proofreading of 20 randomly selected pages of 0&M manual GEK-7577, Grand Gulf 1 and 2, Operation and Maintenance Instructions, Fuel Handling Platform.
Problems identified from the review were as follows:
a.
No ERM existed to document the required review of one manual.
A nonconformance was issued for this item (see item B.4 above).
b.
The issued copy of Manual GEK-75662A did not include three lines on one page examined.
The missing lines were a description of certain automatic fault lock-out conditions for the fuel handling platform operation.
GE procedures require that between initial Operation and Maintenance Instruction Manual distribution and plant turnover, or a date agreed upon by the project of fice, necessary changes will be made to the manuals via errata and addenda.
At the time of plant turnover, or a date agreed upon by the project office, the manual will be reviewed for required corrections anC issued.
Therefore, any errors detected would be corrected during the final issue.
The 0&M manuals are based on approved and verified design documents and contain copies of selected design documents and the 0&M manuals require engineering review and approval.
The 0&M manuals for safety-related equipment are very similar from project to project, therefore, the same technical information receives many reviews during project completion.
It appeared that GE was encountering some problems with typographical errors on recent draft manuals due to typing of an entire manual for the first time on a word processing machine.
However, there was no evidence to indicate that the manuals were being issued with a large number of typographical errors or being issued with known errors.
This item is considered c1] sed with respect to concerns regarding GE issuing O&M manuals with known errors.
81
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION NO.
99900403/82-01 RESULTS:
PAGE 6 of 10 2.
Potential Document Control Problems at GE (La Salle project) -
During a Region III project specific (La Salle) inspection of GE, discrepancies were identified relative to the reference document revision numbers listed in GE Design Report 22A7431, Main Steam Piping and Equipment Loads, and the document revision numbers listed in computer printouts and in the GE Engineering Information System (EIS).
The EIS is a computer listing of GE design documents which shows the latest revision status.
The design report noted was a stress report and the report was certified to the applicable revisions of input documents on December 15, 1981.
The reference documents noted in the design report were revised, but the revisions had no effect on the results.
Three additional design reports were examined and certain reference documents were compared to the EIS computer printout.
No discrepancies were identified.
Since a certification of stress reports is required for applicable input documents, the one case noted where the reference documents list did not indicate the latest revision is not considered a problem.
This item is considered closed.
3.
10 CFR Part 50.55(e) From River Bend Project - This area of inspec-tion resulted from a 10 CFR Part 50.55(e) report from the River Bend project stating that the heat loads provided to S&W by GE for the sizing of HVAC in the HPCI diesel generator room were about one-third the actual values.
An S&W engineer apparently questioned the values of heat radiation provided.
GE had provided S&W with heat load values in a specification for preliminary sizing of HVAC equipment.
The l
diesel generator was supplied by Stewart and Stephenson (S&S) and values of heat radiation were indicated in an operation manual.
When S&W questioned the values provided, GE contacted S&S and an S&S engineer provided a memo with a revised value which was about three times the preliminary GE value and the value provided by the S&S manual.
No explanation was given for the higher value or the original value.
GE is presently evaluating this item as a potentially reportable condition and apparently no final value has been provided to S&W.
Morrison Knudsen also provides HPCI diesel generators to GE and their heat load value was also lower (about one-third to one-half) than the latest value provided by S&S.
GE stated that the followin0 sites could be affected; La Salle, Nine Mile Point 2, Perry, Clinton, River Bend, Hanford, Grand Gulf, Allens Creek, TVA (all), Skagit, and Black Fox.
This item will remain open pending further review at S&W, S&S and GE.
82
ORGANIZATION:
GENERAL ELECTRIC COMPANY l
NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORI IN5PECTION NO..
99900403/82-01 RESULTS:
PAGE 7 of 10 4.
10 CFR Part 50.55(e) From Hartsville Project - This area of inspection resulted from a 10 CFR Part 50.55(e) report for Hartsville and Phipps Bend (TVA) stating that GE has not been advising TVA of GE initiated potentially reportable conditions (PRC's) or of GE or C. F. Braun initiated STRIDE Design Deficiency Reports (SDDR's) and that GE has not been evaluating these PRC's and SDDR's for reportability in accordance with 10 CFR Part 50.55(e).
Upon review of this item, it was noted that an NRC staff meeting was held on December 10, 1981, at the request of GE, to discuss how 10 CFR Part 50.55(e) data is being applied to non-licensee organizations.
GE was advised by the NRC staff that the Construction Permit (CP) holder must assure, through instructions to his contractors, that he is informed of matters reportable under 50.55(e) and that the CP holder and his contractor must decide on whether contract requirements on this issue are being satisfied.
GE's present method is to evaluate both deficiencies and deviations under 10 CFR Part 21 criteria and once a condition is deter-mined not to be reportable under Part 21, the licensee is not notified of the condition.
As a result of the December 10, 1981, meeting, the NRC staff informed GE that additional guidance regarding the subject would be issued, probably in the form of an Information Notice.
GE procedure s E0P 65-4.00 (Potentially Reportable Conditions) and Group Procedure 70-42 (Reporting of Defects and Noncompliance Under 10 CFR Part 21 or Part 50.55(e)) were reviewed and they provide the necessary control for GE to meet their reporting requirements to the NRC.
This item is considered closed with respect to concerns regarding GE reporting requirements to the NRC.
5.
Status of Previously Inspected Items:
a.
Electro Switch, Series 20, Type PR Investigative activity is continuing.
An action item has been assigned to Quality Assurance and identified in an internal letter.
The action item indicates that Quality Assurance should 83
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION I
NO.
99900403/82-01 RESULTS:
PAGE 8 of 10 (1) verify that switches with mirror image contacts had not been shipped to other ust 3, and (2) determine how an improperly configured switch was shipped to the user.
This item remains open and will be evaluated during a future inspection.
b.
Incompatibility of Design Documentation with Shipped Ter-mination Cabinets - The inspector was informed that this deficiency occurred due to lack of control of test assemblies, i.e. changes were made to the test assemblies as a result of changes in cable ladders and termination points during the staging activity.
Currently, test assemblies are subjected to the same design control measures as deliverable hardware.
This practice was instituted prior to complete shipment of hardware for Grand Gulf 1.
The problem appears isolated to the Grand Gulf Nuclear Generating Station.
It should be noted that test assemblies are substitutes for the termination cabinets which are shipped about a year prior to the staging activity at the factory.
The inspector was informed that incorporation of checklists into design record files was a separate and unrelated issue regarding incompatibility of design documentation with shipped termination cabinets.
j A review of ccntractual documents failed to reveal a require-ment for supplying "as shipped" drawings.
Design record files were reviewed for incorporation of checklists.
Check-lists had been incorporated and reflected design verification.
Actions related to this area of the inspection appear adequate.
I This item is closed.
6.
10 CFR Part 50.55(e) Reports From Mississippi Power and Light Company (MP&L Co.).
a.
Concerning inadequate circuit separation in the Power Generation Control Complexes (PGCC) at the Grand Gulf Nuclear Station, Units 1 and 2 (GGNS 1 & 2), the inspector was informed that:
(1) The exact cause of the problem had not been identified; 84
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA RLPORT INSPECIl0N NO.
99900403/82-01 RESULTS:
PAGE 9 of 10 (2) Reg Guide 1.75 and IEEE 384 became effective at or near completion of the design phase of the PGCC for GGNS, Unit 1; and (3) Available talent perceived that the wiring complied with the requirements.
Continuous education and evaluation revealed this perception to be in error.
Fire test specifi-cation, fire test report calculations, and other pertinent data for these analyses and tests had been incorporated into Design Record File No. A00-794-6.
A review of this file and other related information indicated that adequate corrective action had been taken or planned.
During the review, it was noted that a test report did not contain the signature of the performer as required by the test specification.
This was documented as a nonconformance. (See B.1 above). Prior to the close of this inspection, the inspector was provided a copy of a letter that gave an inter-pretation of the headings identified on the test report cover sheet.
Additionally, the test report cover sheet was modified by adding a heading entitled " Performed By."
This heading was completed by the typed name and signature of the performer and the date.
Since corrective and preventive actions were taken, no further response is required.
b.
Concerning incorrect assembly of cable connectors on PGCC cables for GGNS1, the MP&L Co. final report dated June 12, 1981, indicated that the deficiency was not reportable; however, the inspector elected to evaluate actions taken.
The inspector (1) observed preassembly of cable connectors and pin insertion, and (2) reviewed inspection records and special study test data.
While observing pin insertion, it was noted that the work bench contained four disassembled connectors.
The inspector inquired as to the method used for assuring that reassembly was correct and was advised that this practice would be evaluated. While observing connector preassembly it was noted that the appropriate section of the PGCC Cable Assembly Instruction Manual had not been issued to the work station This was documented as a non-conformance (see 0.2 above).
Corrective and preventive actions were taken prior to the close of this inspection; no further response is required.
This item is considered closed.
85
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORI INSPECIl0N NO.
99900403/82-01 RESULTS:
PAGE 10 of 10 c.
Concerning termination of size 8 AWG and smaller wire in junction boxes of PGCC termination cabinets supplied for GGNS 1 and 2, the MP&L Co. final report, dated August 10, 1981, indicated that the deficiency was not reportable; however, the inspector elected to evaluate actions taken.
An unmounted junction box terminal block with attached sizes 8 AWG and 14 AWG wires was observed, as well as an actual installation in a remote shutdown vertical board.
Also, drawings and manufacturing standard practices governing assembly were reviewed.
All appeared in order and this item is considered closed.
d.
Transmitters and trip units for reactor vessel water level scrams supplied for GGNS 1 can not be adjusted to the specified setting.
The Potentially Reportable Condition File, No. 81-23 contained (1) an internal memorandum, dated April 20, 1981, which provided explicit details of the problem, and (2) a memo of telecon, dated August 19, 1981, which announced verbal notification to the Commission.
It was not apparent, from reviewing the file, that further evaluation was underway from April 20 to August 19, 1981; however, it was apparent that the written report to the Commission, dated August 19, 1981, contained no information significantly different from that contained in the internal memorandum, dated April 20, 1981.
The file indicated the following activities were underway from April 20 to August 19, 1981:
(1) identification of the cause of the problem; (2) extent of the problem, and (3) correction of the cause.
Information in the file indicated the problem was the misapplication of equipment caused by "an engineering mistake in choosing the transmitters to be used."
During the exit interview, exception was taken to the NRC inspector's observations regarding the length of time between awareness and reporting of the problem.
It was noted that additional information (requested by the Manager of Safety and Licensing from Quality Assurance),
regarding transmitters and trip units for reactor vessel water level scrams, had not been provided within the time l
required by NEBG Procedure No. 70-42.
This was documented as a nonconformance (see B.3 above).
86
ORGANIZATION:
GIBBS & HILL, INC.
NEW YORK, NEW YORK REPORT INSPECTION INSPECTION NO.
99900524/82-01 DATE(S) 2/22-26/82 ON-SITE HOURS: 26 CORRESPONDENCE ADDRESS:
Gibbs and Hill, Inc.
ATTN:
Mr. P. P. DeRienzo, Vice President Quality Assurance 393 Seventh Avenue New York, New York 10011 ORGANIZATIONAL CONTACT:
Mr. N. N. Keddis, QA Manager TELEPHONE NUMBER:
(212) 760-5450 PRINCIPAL PRODUCT: Architect Engineering and Consulting Services NUCLEAR INDUSTRY ACTIVITY: The total effort committed to domestic nuclear design activities is approximataly 23% of the 2,000 employees of Gibbs & Hill, Inc., at their New York facilities.
Major projects include the design of Comanche Peak, Units 1 and 2; Three Mile Island, Unit 1, FSAR update; Beaver Valley, Unit 1, equipment update, and Bellefonte, Unit 1, design studies.
ASSIGNED INSPECTOR:
G. (_r
- V h /Q - $1 D.F.' fox', Reactor /59stemsSection(RSS)
Date OTHER INSPECTOR (5):
APPROVED BY:
R, YbN1 C. JrCflale, Chief, RSS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and the PSAR and FSAR for the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2.
B.
SCOPE:
This inspection was made as a result of the issuance of a 10 CFR Part 21 report by Gibbs & Hill pertaining to an analysis of the Tornado Venting System for the Comanche Peak Steam Electric Station.
Additional areas included in the inspection were status of previous inspection findings and design inspection of safe shutdown capability following high and moderate energy line breaks outside containment.
PLANT SITE APPLICABILITY:
Comanche Peak Steam Electric Station, Units 1 and 2, Dockets 50-445 and 50-446; Fort Calhoun, Unit 1, Docket 50-285.
87
ORGANIZATION:
GIBBS & HILL, INC.
NEW YORK, NEW YORK REPORI IN5PtLl10r4 NO.:
99900524/82-01 RESULTS:
PAGE 2 of 4 A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
None.
C.
UNRESOLVED ITEMS None.
l O.
STATUS OF PREVIOUS INSPECTION FINDINGS:
(Closed) Nonconformance (81-04):
Contrary to CPSES project procedure DC-5, certain of the changes made in Revision 3 of specification 2323-MS-46A were not identified nor design reviewed, as required.
Specification 2323-MS-46A was revised on February 2,1982, (via DE/CD S-2564) to identify, with line bars and revision numbers, the previously unidentified changes that were made to Revision 3 of the specification.
The revised specification was design reviewed on February 23, 1982.
Procedure DC-5 was being revised to delete the requirement that design reviews be performed on the edited and finished typed document and permit the design review to be performed on a draft version of the document.
The design review would be performed in parallel with the interdis-ciplinary reviews of the document.
The originating discipline would be responsible for assuring that all significant changes made to the draft were design reviewed prior to issue of the document.
Gibbs and Hill management stated that the revised procedure will be issued during the first quarter of 1982.
88
i ORGANIZATION:
GIBBS & HILL, INC NEW YORK, NEW YORK REPORT INSPECTION NO.
99900524/82-01 RESULTS:
PAGE 3 of 4 E.
OTHER FINDINGS OR COMMENTS 1.
10 CFR Part 21 Report Inspection - Review of available docu-mentation, and interviews with cognizant personnel, indicate that the original (February 11, 1977) CPSES tornado venting system pressure and air flow analyses was not sufficiently detailed so as to determine if the pressure differential across interior walls, doors, glass panels, electrical equipment cabinets, etc., would exceed the design allow-able in the event of a design basis tornado (DBT).
The original calculation assumed that the interior of the turbine, auxiliary, safeguards, fuel handing and electrical and controls buildings of the CPSES was divided into approximately 25 " fire zones" and determined the venting area necessary to assure that the maximum differential pressure (DP) between any two contiguous volumes (zones),
or to the atmosphere, did not exceed the design allowable (DA) of 1.5 PSI.
The analysis is currently being redone using the COMPARE computer code.
The affected buildings are assumed to be divided into 95 interior volumes with 163 interconnecting air flow paths.
The status (open and closed) of over 400 interior doors is being considered in the analysis. The preliminary results indicate that approximately 15% of the DPs exceed the DA by over 10%
and that approximately 5% of the DPs exceed the DA by over 40%.
The results of the calculation are being transmitted to the affected disciplines for the evalua-tion of the effect of the increased DPs and air mass flow rates on their designs and air handling systems.
G&H stated that the calculation should be completed by March 26, 1982, and that all required design corrective actions will be completed during 1982.
The inspector noted that the validity of the analyses is dependent upon the status of approximately 246 doors within the identified buildings.
Some of the doors are assumed to be held open (or closed) during a DBT by a locking mechanism.
The status of certain other doors during a DBT must be assured by administrative control.
Further, many of the doors and their associated hardware are being procured as architectural doors and as such are only capable of with-standing a DP of from 0.3 to 0.8 PSI depending upon the direc-tion of the applied loading.
G&H stated that this concern will be addressed during the design corrective action phase of the program.
89
ORGANIZATION:
GIBBS & HILL, INC.
NEW YORK, NEW YORK REPORI INdPtLilVN NO.
99900524/82-01 RESULTS:
PAGE 4 of 4 Unit 1 of the Fort Calhoun nuclear power station was also designed by G&H.
Detailed design records were not available at G&H for examination by the inspector.
G&H management stated that they informed the Omaha Public Power District (OPPD) of the situation prior to October 8, 1981.
Action taken on this item by OPPD could not be determined from the documentation avail-able at G&H.
This item will be followed during future inspections.
2.
Design Inspection - Branch Technical Position Papers and sections of the CPSES Units 1 and 2 FSAR pertaining to postulated piping failures in fluid systems and accident analyses, were reviewed to determine the commitments, design bases, and design measures employed by G&H to assure that the nuclear power station can be safely shut down in the event of the postulated rupture of any high or moderate energy line outside of the reactor containment.
One technical report, 2 analytical models, 22 drawings and figures, 7 calculations, 2 design verification records, and other related documentation were examined by the NRC inspector to verify imple-mentation of commitments.
The NRC inspector determined that some of the subcompartment tempera-ture and pressure analyses performed to date do not appear to support the conclusion that no essential system or component required for safe plant shutdown is rendered incapable of performing its necessary functions as a result of any postulated pipe rupture.
This area will be inspected further during subsequent inspections.
90
ORGANIZATION:
ITT GRINNELL INDUSTRIAL PIPING, INCORPORATED KERNERSVILLE, NORTH CAROLINA REPORT INSPECTION INSPECTION NO.
99900019/82-01 DATE(S) 3/16-18/82 ON-SITE HOURS: 52 CORRESPONDENCE ADDRESS:
ITT Grinnell Industrial Piping, Incorporated ATTN:
Mr.
A.
K.
King President P. O. Box 566 Kernersville, NC 72284 ORGANIZATIONAL CONTACT:
I. E. Johnson IELEPHONE NUMBER:
(919) 996-6400 PRINCIPAL PRODUCT: Nuclear piping subassemblies.
NUCLEAR INDUSTRY ACTIVITY: Not obtained during this inspection.
-m ASSIGNED INSPECTOR: # Do,.#
E ir 2_
(4-H. W. Roberds, Reactive & Component Program Section Date (R& CPS)
OTHER INSPECTOR (S): J. T. Conway, R& CPS APPROVED BY:
.s r2
- 1. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
This inspection was made as a result of: (1) the detection of mechanical enhancement of the specified penetrameter 4-T hole images in Midland, Unit 1 piping weld radiographs; (2) the detection at Midland, Units 1 and 2, of cracks in Type 316 stainless steel piping subassembly welds; and (3) the identification that fittings in piping subassemblies l
furnished to Phipps Bend, Unit 1, and Hartsville, A and B, had been tested (cont. on next page)
PLANT SITE APPLICABILITY:
Penetrameter 4-T hole image enhancement - Docket Nos.
50-518/50-519/50-520/50-521 and 50-329.
Incorrect Fitting NDE Method - Docket Nos. 50-518/50-519/50-520/50-521 and 50-553.
Type 316 stainless steel weld cracks - Docket Nos. 50-329/50-330.
91
ORGANIZATION:
ITT GRINNELL INDUSTRIAL PIPING, INCORPORATED KERNERSVILLE, NORTH CAROLINA REPURI IN5Fttitua NO.
99900019/82-01 RESULTS:
PAGE 2 of 5 SCOPE:
(Cont.) by the ultrasonic method and not by the required liquid penetrant or magnetic particle test method.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section V, Article 2 of the ASME Code, radiography of pipe welds with wall thickness of less than inch was not performed with a technique of sufficient sensitivity to display the penetrameter image and the specified T hole for certain piping assemblies furnished to Midland, Unit 1, and Hartsville, Unit A and B.
A review of radiographs at the Midland site and by NRC personnel at the ITT Grinnell, Kernersville plant, identified that certain radiographs had been mechanically altered to include an artificial representation of the specified 4-T hole of the penetrameter.
2.
Contrary to Criterion V of Apppendix B to 10 CFR Part 50 and Section III of the ASME Code, paragraph NA-4370, documentation was not made available to the NRC inspector which would indicate corrective action requirements had been extended to a subcontractor, who had examined fittings by the ultrasonic method instead of the required liquid penetrant or magnetic particle method.
C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
(0 pen) Deviation (Inspection Report 80-01):.
Copies of forms to be used in resolution of Design Specification conflicts were not being sent to the Director of Quality Assurance, as required by corrective action commitments.
This subject was not addressed during the current inspection.
92
ORGANIZATION:
ITT GRINNELL INDUSTRIAL PIPING, INCORPORATED KERNERSVILLE, NORTH CAROLINA REPORI 1h5Fttilun NO.
99900019/82-01 RESULTS:
PAGE 3 of 5 E.
OTHER FINDINGS OR COMMENTS:
1.
Nuclear Regulatory Commission, Region III, Referral of Penetrameter 4-T Hole Image Enhancement on Radiographs of Piping Welds Fabricated by ITT Grinnell a.
This inspection was performed concurrently with an investigation by the Region IV Investigation and Enforcement Staff.
Investi-gative findings are contained in Report No. 99900019/82-02.
b.
It was discovered that some radiographs of Class 2 pipe welds fabricated by ITT Grinnell (ITT-NC) for Midland, Unit 1, and Hartsville, Units A and B, had been mechanically altered to include an artificial representation of the specified 4-T hole of the penetrameter.
The altered radiographs appear to be limited to pipe that is less than inch wall thickness and where the radiographic technique used required the penetrameter selection be based on the single wall thickness of the pipe.
As the result of an investigation by the NRC Region IV office, it was established that radiographs had been altered, on occasion, for approximately 6 years.
The method of alteration to the radiograph consisted of enhancing the image of the applicable penetrameter T-hole with a lead pencil, thereby, making unacceptable or questionable quality radiographs appear to be acceptable when examined under normal viewing conditions.
2.
Nuclear Regulatory Commission, Region II Referral of Fittings (Incorrect NDE) on Spool Pieces Fabricated by ITT-NC - A detailed review of the documentation (e.g., purchase orders, manuals, pro-cedures, specifications, travelers, nonconformance reports, etc.)
relating to the purchase, fabrication and examination of the forged fittings (both 12" and 20" 0.D.) led to the following findings:
a.
Specific requirements (e.g., chemical, mechanical, NDE, etc.)
were contained in TVA's purchase order to ITT for the spool pieces.
Although ITT-NC referenced the code, three ITT specs, and two ANSI standards in the purchase orders for the fittings to their supplier, ITT-Grinnell of Princeton, Kentucky (ITT-KT),
the specific test and inspection requirements were not addressed.
In addition, there was no evidence that ITT-NC procurement documents required ITT-KT to provide an acceptable quality assurance program.
(These apparent omissions have been corrected on current purchase orders).
93
C RGANIZATION:
ITT GRINNELL INDUSTRIAL PIPING, INCORPORATED KERNERSVILLE, NORTH CAROLINA REPORT INSPECi10N N1 99900019/82-01 RESULTS:
PAGE 4 of 5 b.
QC personnel at ITT-NC perform receipt inspection to verify that the items meet purchase order requirements.
Negative findings result in the generation of a nonconformance report, and the suspect item goes to a segregated storage area to await dis-position.
Contrary to the above, there was no evidence that a nonconforming report had been generated to reflect the fact that a number of fittings were not meeting purchase order requirements in the area of NDE.
There was no evidence that an adequate indoctrination and c.
training pregram was established for personnel performing receipt inspection to assure that they were adept in performing quality affecting activities.
d.
There was no evidence that nonconformance reports are periodically analyzed by the QA organization to show quality trends with the subsequent results reported to upper management for their review and assessment.
i e.
ITT-NC and TVA have reached a mutual agreement as to the disposition of the affected spool pieces containing the noncon-forming fittings.
Some of the spool pieces have been received I
and inspected by ITT in accordance with the original inspection l
requirements.
The remaining spool pieces in question will be sent to ITT for reinspection in the near future.
3.
Midland Type 316 Stainless Steel Piping Weld Cracks - A detailed review of the documentation (e.g., reports, procedures, specifications, travelers, qualification records, etc.) relating to the fabrication and examination of the type 316 stainless steel pipe spool pieces (10" and 12" sizes) containing cracks in the girth welds led to the following findings:
a.
After field surface grinding of welds in preparation for inservice inspection, liquid penetrant examination revealed circumferential linear indications in 20 welds.
The welds were performed by the same individual, and the cracks were confined to a single surface pass made by the submerged metal arc welding process.
b.
The Materials & Quality Services Dept. of Bechtel performed a metallurgical investigation (Ref. report BLN No. 1279-10) of one weld and determir.ed that the linear cracks were due to:
(1) high restraint in the weld pass, (2) low delta ferrite in weld metal; and (3) unfavorable shape of weld bead.
94
ORGANIZATION:
IIT GRINNELL INDUSTRIAL PIPING, INCORPORATED l
KERNERSVILLE, NORTH CAROLINA REPORT INSPECl10N RESULTS:
PAGE 5 of 5 NO.
99900019/82-01 j
c.
ITT's Research Development and Engineering Division concluded (Ref. report no. 2998) that the "
defect indications will not affect the integrity levels applicable to the Class 1 l
and Class 2 piping systems."
d.
NRC inspection of this defect condition did not identify any nonconformance with the welding requirements of Section III of the ASME Code.
No documented eviden:e was made available, however, which would indicate actions had been taken to preclude or minimize occurrence of this type of defect; e.g., changes in weld process controls to reduce base material dilution.
This item will be further reviewed during a future inspection.
e.
All the welds containing linear liquid penetrant indications will be repaired in the field by Babcock & Wilcox using Bechtel procedures, standards, and specifications which were concurred in by ITT-NC.
95
i ORGANIZATION:
ITT GRINNELL, PIPE HANGER DIVISION, ENGINEERING DEPARTMENT j
PROVIDENCE, RHODE ISLAND REPORT INSPECl10N INSPECTION NO.
99900285/82-01 DATE(S) 2/9-12/82 ON-SITE HOURS: 27 CORRESPONDENCE ADDRESS:
ITT Grinnell Pipe Hanger Division ATTN:
Mr. D. M. Sewell Division QA Manager 621 Dana Avenue Warren, OH 44481 ORGANIZATIONAL CONTACT:
Mr. D. M. Sewell, Division QA Manager TELEPHONE NUMBER:
(216) 373-1500 PRINCIPAL PRODUCT: Component Supports NUCLEAR INDUSTRY ACTIVITY: Approximately 65% of ITT Grinnell's work is devoted to the commercial nuclear industry.
T./ '
ASSIGNED INSPECTOR:
(
i
--?
L7 E. Ellershaw, Reactive & Components Program
/Da(e
- Section (R& CPS)
OTHER INSPECTOR (S):
APPROVED BY:
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
i B.
SCOPE:
This inspection was conducted as a result of:
(1) receipt of a potential construction deficiency report (CDR) from Southern California 1
Edison Company pertaining to the use of washers other than hardened steel l
washers; (2) a CDR from TVA pertaining to linear indications being found in I
hanger material; (3) two CDRs from Duke Power Company pertaining to the PLANT SITE APPLICABILITY:
Docket Nos:
50-361; 50-362; 50-438; 50-439; 50-413; 50-267.
97
ORGANIZATION:
ITT GRINNELL PIPE HANGER DIVISION ENGINEERING DEPARTMENT PROVIDENCE, RHODE ISLAND REPORI INSPECIl0N NO.
99900285/82-01 RESULTS:
PAGE 2 of 10 SCOPE: (cont.) use of a riser clamp fabricated by an unapproved vendor and a snubber extension piece being welded off center; (4) a regional request related to incorrect torque requirements resulting in snubber threads being stripped; and (5) items requiring further inspection identified during our inspection at ITT Grinnell, Warren, Ohio.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and QA Manual Section 4.2, verification in accordance with the requirements of product drawings was not performed, in that the product drawings specified the use of carbon steel washers in Mechanical Shock Suppressors, Figures 306 and 307, but brass washers were actually used.
C.
UNRESOLVE ' ITEMS:
ITT has used a material for rivet manufacturing which has not been approved by the ASME Code.
Approval is pending (see paragraph D.1.).
D.
OTHER FINDINGS OR COMMENTS:
1.
Follow Up Inspection Item - The use of a riveted construction for the ITT PE-41 Series Pipe Clamp rather than a welded or bolted construction, was an item that had been inspected at ITT's Warren, 0 hic, facility (Vendor Program Branch Inspection Report 99900282/81-01).
However, certain data could not be substantiated at that time, insofar as this data is maintained at ITT, Providence, Rhode Island, and would be reviewed during an inspection at that facility.
ASME Code,Section III, Subsection NF, imposes mandatory Appendix XVII for the analysis procedure to be followed, unless design is accomplished by use of either the experimental stress analysis procedure, or the load rating procedure.
If Appendix XVII is used, the supports are required to be of a welded, bolted, or welded and bolted construction.
Fabrication of component supports is also required to comply with the provisions of Ar'icle NF-4000, which addresses welded and bolted construction only.
93
ORGANIZATION:
ITT GRINNELL PIPE HANGER DIVISION ENGINEERING DEPARTMF.NT PROVIDFNCE, RH0DE ISLAND REPORI INSPELIlON NO.
99900285/82-01 RESULTS:
PAGE 3 of 10 The inspector reviewed the Load Capacity Data Sheets (LCDS), Revision 0, dated June 1, 1980, and Revision 1, dated June 9, 1980.
Revision 0 did not list the rivets and showed that design had been accomplished by analysis.
In addition, it did not list the required Level A and B, Level C and Level D maximum load pounds.
Revision 1 added the levels and maximum load pounds, but still did not address rivets, and continued to show use of linear analysis.
Review of LCDS, Revision 2 dated January 30, 1981 showed that ITT used the load rating procedure in lieu of analysis, and that the ASME Code classification is Class 1, 2, 3, and MC.
The LCDS identifies the rivet material as being SA 453 Grade 660, Condition A or B.
Revision 3 and 4 of the LCDs, dated August 1 and October 29, 1981, respectively, both reflect a change in material designation for the rivets to AMS 5737.
This is asterisked, with a note stating, "Pending approval of Code Case N-71-11 and N-249-2."
At the time of this inspection, approval had not been obtained from ASME for the use of this material; therefore, this item is considered unresolved.
The following is a brief synopsis of how this problem occurred.
All rivet material was ordered to SA 453 Grade 660, which is a bolting specification.
The drawings provided to the rivet manufacturer also state SA 453 Grade 660.
In addition, the drawings require dimensions tobeinaccordancewitgNAS1199,aNationalAerospaceStandard, titled " Rivet-Solid-100 Flush Head A286 Corrosion Resistant Steel."
NAS 1199 and SA 453 both specify a solution and precipitation heat treatment, but the times at temperature are different.
The rivet manufacturer used the heat treatment specified by NAS 1199, because it is a rivet specification.
The rivet manuf acturer's material test reports show the material as being SA 453 Grade 660, A 286 SS Rivets, Annealed.
lhey also reference the part number as being NAS 1199.
2.
Potential 10 CFR Part 50.55(e) Construction Deficiency Report Southern California Edison Company made a notification to the NRC that washers used on ITT's snubbers are not hardened, in that they are brass and not carbon steel.
99
ORGANIZATION:
ITT GRINNELL PIPE HANGER DIVISION ENGINEERING DEPARTMENT PROVIDENCE RHODE ISLAND RLPORI INSPLLl10N NO..
99900285/82-01 RESULTS:
PAGE 4 fo 10 ASME Code,Section III, NF 4724 states in part, " Bolts tightened by means of a calibrated wrench shall be installed with a hardened washer under the nut or bolt head ITT was notified by Bechtel through issuance of Quality Surveillance Deficiency Report Nos. 81-01 and 81-02, both dated April 3,1981, of this condition.
ITT evaluated the problem by performing hardness tests on the two types of washer material and conducting vibration testing on one of each size snubber assembly involved.
The vibration tests showed that with the use of either the brass or steel washer, the bolted connections performed adequately.
The hardness tests showed that the brass washers were actually harder than the steel washers as they were produced from a cold-worked brass strip material.
The actual hardness readings obtained were Rockwell A 30.9 and 34.3 for the steel washers, and Rockwell superficial 15T 88.7 for the brass.
There are no specific hardness requirements for " hardened" washers.
A review of the associated drawings, BH 1258 and BH 1264, showed that prior to June 4, 1981, the washer standard to be used was Standard C-4-4.1, dated March 21, 1969.
This standard states,
" material: Carbon steel." A revision to both drawings on June 4, 1981, changed the requirement from standard C-4-4.1 to ASTM F436, paragraph 5.
Paragraoh 5 requires washers to have a hardness of Rockwell 38 to 45, except that when hot-dipped galvanized they shall have a hardness of Rockwell C 26 to 45.
The brass washers that were used by ITT are defined in Standard C-4-4.4 dated April 2, 1969, which states with respect to material,
" ASTM B36, alloy 260 or alloy 268."
ASTM B36 is the standard specification for brass plate, sheet, strip, and rolled bar.
The hardness range for the category that was used by ITT is Rockwell B 44-77.
However, brass was not spec ~ified on the drawings and as a result, nonconformance B.1 vas identified.
100
ORGANIZATION:
ITT GRINNELL PIPE HANGER DIVISION ENGINEERING DEPARTMENT PROVIDFNCF. RHODE IStAND REPORT IN5PECI10N NO.
99900285/82-01 RESULTS:
PAGE 5 of 10 3.
Unresolved Item Identified at ITT, Warren, Ohio, Inspection Report 99900282/81-01:
An unresolved item was identified during an inspection conducted at ITT's Warren, Ohio, facility which pertained to Procedure PE-217-1,
" Installation Instructions-ITT Grinnell Pipe Hangers," not addressing the use of spacers at the rear bracket end of sway struts.
If the spacers are not installed, the spherical ball bearing may become completely disengaged for certain sized struts.
ITT revised Procedure PE-217-1 and issued Revision 2 on March 2, 1981, with subsequent Revision 3 being issued on October 16, 1981.
Section V " sway struts," paragraph 2.b states, " Attach the struc-tural attachment insuring that the spacers provided with the assembly are properly installed.
There must be one spacer on each side of the rod end assembly to insure that the ball bushing will remain centered on the load pin so that the full 5 swing can be obtained."
ITT distributed the revised procedure to the following customers on the indicated date:
TVA - March 11, 1981; Texas Utilities Services, Inc. - March 11, 1981; Stone & Webster Engineering Corporation - (Nine Mile Point) -
April 28, 1981; and Associated Piping and Engineering Company (WPPSS 3 and 5) -
February 10, 1982.
This item is considered resolved and will also be noted in the inspection report following the next scheduled inspection at ITT Grinnell, Warren, Ohio.
4.
10 CFR Part 50.55(e) Construction Deficiency Report:
TVA made a notification to the NRC that two examples of hanger material were found to have linear indications at the Bellefonte Nuclear Power Station.
These indications were discovered in areas 1 01
ORGANIZATION:
ITT GRINNELL PIPE HANGER DIVISION ENGINEERING DEPARTMENT PROVIDFNCE, RHODE ISLAND HLPORI IN5PECl10N NO.
99900285/82-01 RESULTS:
PAGE 6 of 10 adjacent to welds which were being liquid penetrant examined.
The size of the indications were such that they were not detected during final inspection by visual means.
Iff did perform an evaluation, by taking cross sections of the affected dreds and performing photomacro and micrographs.
The results showeo indications up to 0.017" depth and 0.045" in length.
The evaluation was reported in Report No. 3057 dated August 28, 1981, " Resolution of Surface Indications In M Beams of ASME SA-36 Steel, Bellefonte Nuclear Power Station, Tennessee Valley Authority."
The results of the evaluation concluded that the linear type indications were surface lap conditions which occassionally occur in structual steels dnd that they do not affect the integrity of the beams, thus are con-sidered acceptable for the service intended.
5.
10 CFR Part 50.55(e) Construction Deficiency Report:
Duke Power Company made a notification to the NRC that riser clamps had been received at Catawba Nuclear Station, Unit 1 which had been fabricated at ITT's Henderson, Tennessee facility, an unapproved source for providing components to nuclear sites.
One Figure 261, 6" riser clamp was identified and returned to ITT for replacement.
The Henderson facility provides commercial items to the Warren plant for use in non-nuclear component supports.
The Warren plant maintains an inventory of these parts and it appears that when a bulk shipment of riser clamps was made to Duke Power, one clamp was used from the Henderson supplied inventory.
ITT Quality Control personnel did not detect this clamp prior to shipment, even though there is a unique identification on Henderson supplied material.
This item is considered an isolated incident.
ITT's Quality Control Manager has held training sessions with all inspection personnel per-taining to material identity and control, and the correct usage of commercial and nuclear components.
102
ORGANIZATION:
ITT GRINNELL PIPE HANGER ENGINEERING DEPARTMENT PROVIDENCE, RHODE ISLAND REPORI INSPLC110N NO.-
93900285/82-01 RESULTS:
PAGE 7 of 10 6.
Potential 10 CFR Part 50.55(e) Construction Deficiency Report:
Duke Power Comapny made a notification to the NRC that a snubber had been received at the Catawba site in which the 6" extension piece had been welded 1/4" off center.
Duke Power Company returned the snubber extension piece back to ITT l
Warren with nonconformance report number 13635, dated December 28, l
1981.
This piece was replaced; however, ITT performed an analysis considered ultra conservative which included the worst-case mismatch and offset variables.
ITT's conclusion, noted in Report No. SA-2824 dated January 21, 1982, showed that the Figure 207N, size No. 1, 4" stroke snubber with the extension piece, would not meet the rated load for the Level C condition (2010.81 lbs. vs. 2067 lbs).
The report further stated the " qualified load is approximately 3% less than the rated load (2067 lbs).
The BEAMCL computer program is a conservative analysis, therefore, 3% difference in rated load is justifiable for this analysis.
The extension piece for Fig. 307N size No. 1, 4" stroke is qualified for the rated load with pin-to pin distance of 59\\" max."
Discussion with the Warren QC Manager indicated that fixturing/
templates are used when welding the extension piece to the plate and the item is inspected.
The QC Manager committed to a thorough review of the welding / inspection process for these items.
This item will require followup during the next inspection at ITT Grinnell, Warren, Ohio.
7.
Follow Up On Regional Request:
NRC Region IV requested a followup inspection at ITT Grinnell, predicated on a nonconformance report and action request by Public Service Company of Colorado (PSC) pertaining to torque specifications which resulted in causing snubber threads to strip at Fort St. Vrain.
The request indicated that the problem was associated with the 1 " cylinder "K" type, which is the only size that Fort St. Vrain has utilized and concern was expressed that ITT may not have furnished manufacturer's recommendations concerning the torque limits for the "K" type snubber to other sites.
103
ORGANIZATION:
ITT GRINELL PIPE HANGER DIVISION ENGINEERING DEPARTMENT PROVIDENCE, RHODE ISLAND kEPORI lhbPELi10N i
NO.
9990C285/82-01 RESULTS:
PAGE 8 of 10 PSC's Nonconformance Report (NCR) No. 81-13, dated February 3,1981, states that the torque value for tie rods is too high, causing threads to strip and the ITT Grinnell repair manaul calls for 21 ft. lbs. for 1\\" Miller type snubbers.
It further states that the cy!inder tie rods for ITT's type "K" Miller cylinder should be torqued to 711 ft. lbs. per the manufacturer's (Miller) instructions.
It also identified the tie rods as being 1/4".
The NRC inspector determined that ITT has supplied approximately 580 safety-related snubbers (Miller cylinders), which consisted of 1 " and 2" "J" type, and 1\\", 2 ", 3h", 4" and 5" "H" type.
ITT's procedure No. PHD-6511-6 dated December 14, 1976, Section H, paragraph 13, states, " Install tie rod hex nuts and torque evenly to requirements shown below:
H Type Cylinder-Size Torque (ft/lbs.)
1" 21 2"
54 3
95 4"
150 5"
315 6"
465 A review of the Miller Fluid Power Company's (MFP) catalogue and a telephone call to MFP on February 11, 1982, revealed that the only type cylinders manufactured have been types A, H, and J.
The MFP catalogue further shows that the " tie rods (discussed in PSC's NCR) are used with the 1 " "J" type cylinder, and the torque value is stated as being 8 ft/lbs.
l l
(
104
ORGANIZATION:
ITT GRINNELL PIPE HANGER DIVISION ENGINEERING DEPARTMENT PROVIDENCE, RHODE ISLAND REPORI INdPLC110N NO.
99900285/82-01 RESULTS:
PAGE 9 of 10 It would appear that personnel at Fort St. Vrain applied the specified torque of the 1 " "H" type cylinder (21 ft/lbs.) to the 1 "
"J" type cylinder (8 ft/lbs.) which causes stripping of the threads.
8.
Unresolved item Identified at ITT, Warren, Ohio, Inspection Report 99900282/81-01:
An unresolved item was identified during an inspection at ITT's Warren, Ohio facility, pertaining to deficient welding symbology used in ITT drawings, as reported by TVA in a CDR regarding Bellefonte, Units 1 and 2.
ITT designated the use of fillet welds for skewed T-joints having angles less than 60 or more than 135 AWS D1.1 requires groove welds for these configurations.
ITT deleted the parenthesis around dimensions in the welding symbols on certain detail sheets resulting in depth of groove weld preparation being indicated rather than the intended groove weld throat size.
Prior to February 1, 1979, all welds were designated as fillet welds by ITT.
An informal document titled "Everything You Ever Wanted to Know About Fillet Welds and Were Afraid to Ask!" was developed and issued on February 1, 1979.
This document was used as a training aid for the engineering department personnel.
This document addressed the requirements of AWS Dl.1 in addition to ASME Code requirements.
As a result, the introduction of groove welds in angles less than 60 or greater than 135 was made, and all drawings developed after that date reflect groove welds rather than fillet welds.
TVA had accepted numerous drawings (pre-1979) which showed fillet welds.
The problem arose when TVA received drawings which were developed after February 1979, and which showed groove welds.
The contract documents between TVA and ITT reference both ASME Code Section IX and AWS D1.1.
ITT's design calculations determine a required load per linear inch based upon joint loading and l
l 105
ORGANIZATION:
ITT GRINNELL PIPE HANGE1 DIVIS10ii ENGINEERINi: DEPARTMENT PROVTOFNCF. RH00E ISLAND REPORT INSPECTION NO.
99900285/82-01 RESULTS:
PAGE 10 of 10 geometry.
These values are compared with Engineering Standard ES-16, " Weld Detign Procedure For Component Supports," which is applicable for w alded connections under the jurisdiction of the ASME Code.
Allasable loads per linear inch were calculated for the various til et weld sizes and it was determined that the weld metal allowable load based on the calculated effective throat always exceeded the allowable loads stated in ES-16, thus the fillet welds mest ASME Code requirements.
ITT's policy pertaining to the use of parenthesis in bevel groove welds symbols, s that dimensions refer to effective throat and rot depth of prepari tion, regardless of whether or not the parenthesis is omitted.
In addition, the tail of the symbols show, "no machining required,"
thus there woulc not be a depth of preparation.
Further, these welds are filed welds and AWS 01.1 states that the engineer (designer) cannot specify the " depth of groove" without knowing the welding process and/or position of welding.
AWS D1.1 is explicit in stipulating that only the effective throat is to be specified on design drawings for partial joint penetration groove welds.
Therefore, these items are considered resolved and will also be noted in the inspection report following the next scheduled inspection at ITT Grinnell, Warren, Ohio.
1 1
106 l
ORGANIZATION:
JOSEPH OAT CORPORATION CA'iDEN, NEW JERSEY REPORi IN5PECi10N INSPECTION NO.
99900251/02-01 DATE(S) 3/8-12/82 ON-SITE HOURS: 31 CORRESPONDENCE ADDRESS:
Joseph Oat Corporation ATTN:
Mr. J. Benckert, Manager Quality Control 2500 Broadway Camden, NJ 08104 ORGANIZATIONAL CONTACT:
Mr. J. Benckert, Manager, Quality Control TELEPHONE NUMBER:
(609) 541-2900 PRINCIPAL PRODUCT:
Nuclear Heat Exchangers and Fuel Storage Racks NUCLEAR IN0llSTRY ACTIVITY:
Approximately 75% of Joc".ph Oat Corporation's activity is devoted to the commercial nuclear industry.
ASSIGNED INSJECTOR:
Y/
W E
2-Ro'ss L. Brown, Reactive & Components Program
~Date Section (R& CPS)
OTHER INSPECTOR (S):
\\
/ nm APPROVED BY:
/.
6' * -.
+ // r h L
- 1. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B; and 10 CFR Part 21.
B.
SCOPE:
Management meeting and status of previous inspection findings.
PLANT SITE APPLICABILITY:
Not identified.
n l
107 L
ORGANIZATION:
JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY RLPORI IN5PECTION NO..
99900251/82-01 RESULTS:
PAGE 2 of 4 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Violation (81-02):
The inspector verified that Section 206 of the Energy Reorganization Act of 1974 has been posted on the employee bulle in boards in the production shop.
2.
(Cl sed) Nonconformance A.1 (81-02):
Review of weld and heat record fr.c Job 2349-M verified that the record was corrected by QC af ter per-formance of a review of weld withdrawal slips which identified the weld seam number, weld procedure, welder, and weld electrode heat number.
Indoctrination and training record dated June 26, 1981, verified that the plant manager conducted a training session with all welding personnel relative to the recording of pertinent information for all weld seams.
3.
(Closed) Nonconformance A.2 (81-02):
Review of the weld and heat record sheets verified that the applicable information was recorded for tube to tube sheet welds identified on drawing 6125, Revision 3, and 6128, Revision 4.
I Joseph Oat Corporation (OAT) letter to their customer, dated August 28, 1982, requested that they file the identified filler metal records in the job pacakges for Job Nos. J-2339-A through M.
4.
(Closed) Nonconformance B (81-02):
Review of the interoffice mamorandum verified that written instructions were issued to the storeroom personnel relative to the disposition of returned electrode.
108
ORGANIZATION:
JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY 4
RLP0kl ikbFtLi1Uh NO.
99900251/82-01 RESULTS:
PAGE 3 of 4 5.
(Closed) Nonconformance C (81-02):
The welder training session held on June 26, 1981, instructed the welders on the need for record accuracy.
The QC daily inspection record verified conformance with the requirements.
6.
(Closed) Nonconf ormance D.1 (81-02):
Operation No. 4 on Master Traveler for Job No. 2441-2 has been revised to include the partition plate (longitudinal baffle) to shell welds as detailed on drawings D-7015-5 and D-7014-2.
7.
(Closed) Nonconformance D.2 (81-02):
Review of the weld heat record sheets for Job No. 2441-2A and B verified the records have been revised to include the applicable information for weld No. 146.
8.
(Closed) Nonconf ormance E (81-02):
Review of the applicable weld procedure (WPS-22, Rev. 12) verified that backgouging and grinding the backside of the weld to sound metal to remove the tack welds is required.
The review of three travelers verified the removal operation had been performed, inspected, and approved.
Review of four fitter welder records verified they are qualified per ASME Code Section IX.
9.
(Closed) Nonconformance F (81-02):
Review of drawing D-7015, Rev. 5, and D-7013 travelers for Job 2441-28, and weld and heat record for this job verified that the drawing had been revised to identify the weld and that the appropriate activity iias been accomplished and information documented.
10.
(Closed) Nonconformance G (81-02):
The inspector verified that the discrepancy in test results of the 7018 electrode used on Job 2335 for the nozzle welds in three Reaction Products Separation Tanks (RPS) i for the Clinch River Breeder Reactor Plant (CRBRP) was transmitted to the customer (General Electric Company (GE)) in the Vendor Case Record VCR-J-2335-44.
GE acceptance of the deficiency was verified in their letter XL-590-10487 which stated GE has also reviewed the stress in the nozzle where the weld wire was used and found that the weld is lightly loaded, and it is GE's conclusion that the test discrepancy in weld wire has no adverse effect on the safety or function of the tanks.
109
ORGANIZATION:
JOSEPH DAT CORPORATION CAMDEN, NEW JERSEY REPORT in 3,'t L i i ur4 NO.
99900251/82-01 RESULT 5:
PAGE 4 of 4 11.
(Closed) Nonconformance H (81-02):
Review of the OAT forming quali-fication procedure, five laboratory reports representing at least three heats of material for the RPS tanks at CRBRP project and GE's letter SL-590-10487 verified acceptance of Gat forming procedure.
Oat did not attempt to universally qualify the tank material but rgther qualified the forming procedure at the specified temperature of 10 F.
110
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION INSPECTION NO.
99900100/82-02 DATE(S) 1/27-28/82 ON-SITE HOURS: 12 CORRESPONDENCE ADDRESS:
Limitorque Corporation ATTN:
Mr. T. Mignogna President 5114 Woodall Rd.
Lynchburg, VA 24506 ORGANIZATIONAL CONTACT:
Mr. K. Groome, Quality Control Manager TELEPHONE NUMBER:
(804) 528-4400 PRINCIPAL PRODUCT: Electric Motor Actuated Operators NUCLEAR INDUSTRY ACTIVITY: Limitorque Corporation (LC) supplies electric motor actuated operators for valve operation.
LC nuclear involvement represents approximately 10 percent of their total production.
'./
ASSIGNED INSPECTOR:
7 '
- Wm. D. Kelley, Reactive & Component Program Section Date (R& CPS)
OTHER INSPECTOR (S):
C' -
APPROVED BY:
~
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
This inspection was made as a result of: (1) Construction deficiency reports by: (a) Consumers Pow r Company concerning the installation of under-rated terminal blocks on component cooling water and service water system electric motor actuated valve operators at the Midland Nuclear Power Plant, Units 1 and 2; and (b) Louisiana Power and Light Company concerning the furnishing to Waterford Generating Station, Unit 3, of valve operators which (SCOPE cont. on next page)
PLANT SITE APPLICABILITY:
Docket Nos. 50-329; 50-330; 50-382: 50-361; and 50-362.
111
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORI IN3FttiiUN NO.
99900100/82-02 RESULTS:
PAGE 2 of 8 SCOPE: (Cont.). contained motor pinion keys that had been made from a wrong material; and (2) a 10 CFR Part 21 report by Bechtel Power Corporation concerning the furnishing to Southern California Edison San Onofre Nuclear Generating Station, Units 2 and 3, of original and replacement motor pinion keys which had been made from a wrong material.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion II of Appendix B to 10 CFR Part 50 the quality assurance program did not provide for indoctrination and training of personnel performing activities affecting quality, other than those personnel reporting directly to the Quality Control Manager; i. e.,
training was not addressed for engineers, parts and shipping personnel, purchasing agents, order processors, and field service personnel associated with nuclear orders.
2.
Cont ary to Criterion V of Appendix 8 to 10 CFR Part 50 and paragraph A.2 of Section IV of the Quality Assurance Manual, a telephone order for five " Keys - Motor Pinion Gear," Drawing Number 60-563-0154-1, Revision A, was received and the parts shipped without the order being routed through the File Department.
3.
Contrary to Criterion XVI of Appendix B to 10 CFR Part 50, the cause and corrective action to preclude recurrence were not documented in regard to:
Furnishing of incorrect carbon steel keys with Certificates of a.
Compliance indicating them to be 4140 alloy steel; b.
Installation of improper terminal blocks by a Field Service Engineer.
4.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph D.la of Section IX of the Quality Assurance Manual, the Shipping Department Supervisor did not ensure that all material shipped to a customer was properly identified, as evidenced by the shipment of 1030 carbon steel keys for an order (WKM Valve Division) requiring 4140 alloy steel keys.
112
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA RLPORT INSPECIION NO.
99900100/82-02 RESULTS:
PAGE 3 of 8 C.
UNRESOLVED ITEMS:
None D.
OTHER FINDINGS OR COMMENTS:
1.
Construction Deficiency Report by Louisiana Pcwer and Light Company (LPLC) - Problem reported was the discovery, as a result of the LPLC follow up of NRC IE Information Notice 81-08, that 10 LC valve operators furnished to Waterford Generating Station had keys of the wrong material holding the pinion gear to the electric motor shaft.
a.
Background
NRC IE Information Notice 81-08, dated March 20, 1981, was issued as a result of the failures of the key holding the pinion gear on the motor shaft of the LC valve operators at the following plants:
(1) Nebraska Public Power District - Cooper Nuclear Station (2) Boston Edison Company - Pilgrim Nuclear Power Station, Unit 1 (3) Georgia Power Company - E. I. Hatch Nuclear Plant, Units 1 & 2 (4) Power Authority of the State of New York - James A. Fitz-patrick Nuclear Power Plant.
LC responded to the notice by notifying all of their customers by letter dated June 3, 1981, of their recommendation to replace all 1018 keys on their Size SMB-3 and SMB-4 valve operators equipped with motors that developed 150 foot pounds torque or greater, and all size SMB-5 valve operators.
LC's letter stated that the replacement keys and their material certificate of compliance would be provided at no cost.
113
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORI IN5Pttilua NO.
99900100/82-02 RESULTS:
PAGE 4 of 8 b.
Findings LC informed NRC IE:HQS in their letter of June 9, 1981, that they had reviewed their design calculations and had found the stresses in the keys of the 1018 material to be acceptable for normal valve operator service; however, the use of a 4140 key increased the ability of the valve operators to withstand impact and shock loads imposed by frequent reversal of the valve travel and/or the stalled motor condition.
LC stated verbally it was their engineering judgment that the key failures were aggravated by the operation of valves without the flowing media in the pipe lines, sudden reversal of the valve travel, and/or electrically jumping the torque switch to effect a tight shutoff of leaking valves.
As stated in LC letter of June 3, 1981, to their customers, no records were available prior to May 25, 1981, that would identify which key material was used on a particular order.
Prior to this date LC had purchased their keys in bulk lots and performed inventory by weight count once a year.
The inspector verified that LPLC had been notified by LC of their recommendation to replace the keys of 1018 material with keys of 4140 material.
2.
Construction Deficiency Report by Southern California Edison (SCE) -
Problem reported was the discovery that the keys purchased from the WKM Valve Division of ACF Industries, Inc., for holding the pinion gear on the valve operator motor shaf t, were 1030 carbon i
steel material and not the 4140 alloy steel indicated by the certificate of compliance.
a.
Background
Bechtel Power Corporation (BPC) reported to NRC RIV by letter dated November 29, 1981, of improper key material in LC valve i
114
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRG!NIA HEPORI lhdPtul10h NO.
99900100/82-02 RESULTS:
PAGE 5 of 8 operators installed on WKM valves at the SCE-San Onofre Nuclear Generating Station.
Replacement keys of 4140 material were ordered from WKM who in turn ordered the keys from LC by TWX dated June 4, 1981.
The NRC RIV inspector was informed by LC personnel that these keys were subsequently lost in the mail between WKM and San Onofre Nuclear Generating Station WKM requested by telephone for LC to ship replacement keys and confirmed the order with a telecopy of their purchase order on July 21, 1981.
The keys were withdrawn from stock and (with an accompanying material certificate of compliance) were shipped on August 19, 1981.
Prior to installation, the replacement key material was analyzed by an unnamed independent laboratory who determined the key material to be 1030 carben steel.
b.
Findings The NRC inspector observed that the keys of 4140 material are identified by a part number which is affixed to the front of the storage bin.
The bin was observed to be adjacent to a bin containing keys of identical size of 1018 material.
The keys do not have a material identification marking and keys of both materials were available to the employee withdrawing the keys for the WKM replacement order.
The Manager of Quality Control issued a material certi-ficate of compliance stating the keys were 4140 material.
The NRC inspector reviewed the limited number of documents available pertaining to the keys which consisted of the meterial certificate of compliance, the part drawing, a shipping list and an acknowledgment of the purchase order.
The NRC inspector requested to see a copy of the noncon-formance and corrective action documentation, and was informed corrective action had been implemented, but neither the nonconfort.ance and cor: 'ctive action, nor the retraining of the employee had been documented.
Three nonconformances were identified.
c.
Items Requiring Follow-up Inspection - The LC Material Review Board did not meet and decide the required corrective action with respect to identified cause of furnishing incorrect key materials to nuclear customers; i.e., material 115
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPuRT itortunion
}
NO.
99900100/82-02 RESULTS:
PAGE 6 of 6 identified as 1030 steel and 1018 steel were shipped to SCLC, San Onofre Nuclear Generating Station and LPLC, Waterford Generating Station repectively, as 4140 steel.
A detailed evaluation of the LC quality assurance program provisions for material control will be.made at a subsequent inspection, in order to assure compliance with Criterion VIII of Appendix 8 to 10 CFR Part 50.
3.
Construction Deficiency Report by Consumer Power Company (CPC) -
Problem reported was the discovery of underrated terminal blocks in LC valve operators that had been furnished by valve manufacturers to the Midland Nuclear Power Plant.
a.
Background
CPC-Midland Nuclear Power Plant notified Region III of the NRC by letter dated February 13, 1981, of a defective item reportable under paragraph 50.55(e) of 10 CFR Part 50.
The attachment to the letter states in part, "
The other 10 operators inspected had Marathon Series 300 blocks which were initially rated by Marathon for 600-Volt Service and have subsequently been derated to 300 volts by their catalog 10M79 b.
Findings LC stated in their letter to the Henry Pratt Co., dated January 21, 1981, that Marathon Special Products (MSP) had an Under-writer's Laboratory (UL) approval on the MSP 300 Series terminal I
block which specified that the terminal block rating was based on dimensional requirements and not on breakdown voltage.
Currently the MSP 1979 catalog shows the 300 Series terminal block as carrying a UL rating of 300 volts.
The LC letter referenced an article in the April 7, 1977 issue of Machine Design which states that most manufacturers rate their terminal blocks at one third of the breakdown voltage.
The NRC inspector verified that the MSP rating sheet for their 300 Series terminal block states the line to line breakdown voltage is 9,000 volts and the line to ground breakdown voltage is l
116
ORGANIZATION:
LIMIT 0PQUE CORPORATION LYNCHBURG, VIRGINIA REPORT IN5PECTION MO.
99900100/82-02 RESULTS:
PAGE 7 of 8 11,200 volts.
It is LC's position based on past experience and the actual breakdown voltage that the MSP 300 series terminal blocks are suitable for service in their valve operators because the motors operate on 460 volts.
4.
Construction Deficiency Report by Consumer Power Company (CPC) -
Problem reported was the discovery of terminal blocks for 120 volt service installed on LC valve operators for 460 volt service at the Midland Nuclear Power Plant.
a.
Background
CPC-Midland Nuclear Power Plant notified Region III of the U.S. Nuclear Regulatory Commission by letter dated February 13, 1981, of a defective item reportable under para-graph 50.55(e) of 10 CFR Part 50.
The letter states an in.estigation revealed that eight out of 18 LC valve operators had MSP Series 100, Clinch Jones Series 140, or Beau Products terminal blocks, which are underrated for 460 volt service.
b.
Findings A newly hired LC Field Service Engineer from the Willowbrook, Illinois, Service Center arrived at the CPC Midland Nuclear Power Plant on November 21, 1977, to make architect-engineer requested field changes of the terminal blocks on LC valve operators.
The change requested was the replacement of five
)
terminal blocks on 71 valve operators. The architect-engineer had requested the change to permit one vacant terrinal between the three 460 volt terminals connected to the motor and the two 120 volt terminals connected to the motor heater.
Some of the motorized valve operators were modified at the Henry Pratt Company.
The LC Lynchburg plant shipped 66 MSP 300 Series I
terminal blocks to the LC Willowbrook Service Center on f
November 16, 1977.
Some of the terminal blocks were damaged in shipment, so an additional 20 Clinch Jones Series 140 terminal blocks were shipped on November 28, 1977; however, these terminal blocks could only be used in 120 volt control circuits and not the 460 volt power circuits.
LC stated in their letter of January 21, 1981, to CPC that, "The field service engineer that performed the required terminal strip changes was a new employee and did not know at the time that the Clinch Jones terminal strips were not adequate to be used for power connections."
117 j
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT irdw t.L i t ura NO.
99900100/82-02 RESULTS:
PAGE 8 of 8 At LC's request terminal blocks were removed from several installed valve operators, and in addition to finding Clinch Jones Series 140 terminal blocks installed, a terminal block which appeared to be an MSP Series 100 was found installed on a valve operator supplied on LC Grder 3A2337-B.
LC examined their records and could not find any record where they had purchased MSP 100 Series terminal blocks, and therefore, concluded that it was possible that the inex-perienced LC field service engineer obtained this terminal block locally at the job site.
The inspector reviewed one LC letter, one service call report, two lists of LC motorized valve operators, two shipping lists, one customer order acknowledgement and one shipping memo.
One nonconformance was identified.
Item Requiring Follow up Inspection - Section 1 of the LC Quality c.
Assurance Manual states that it is the intent of the manual with its supporting procedures to meet the applicable provision of Appendix B to 10 CFR Part 50.
Criterion I of Appendix B requires the individuals assigned the responsibility for assuring effective execution of any portion of the quality assurance program to have direct access to such levels of management as necessary to perform this function.
The Quality Assurance Manual splits the authority and responsibility for the imple-mentation and enforcement of the quality assurance program between the Quality Assurance Administration, who reports to the j
Vice President Admimistration, and the Quality Control Manager, who reports to the Plant Manager.
There are no documented lines of communication between the Quality Assurance Administration and the Quality Control Manager.
A detailed evaluation of the LC organization will be made at a subsequent inspection, in order to assure compliance with Criterion I of Appendix B to 10 CFR Part 50, with respect to plant quality personnel having required authority and organizational freedom, 118
ORGANIZATION:
PACIFIC AIR PRODUCTS COMPANY SANTA ANA, CALIFORNIA REf0Ri IN$PEC110N INSPECIION NO.
99900769/82-01 DATE(S) 2/8-11/82 ON-SITE HOURS:
27 CORRESPONDENCE ADDRESS:
Pacific Air Products Company ATTN:
L. R. Hess, President 3133 Harvard Blvd.
Santa Ana, California ORGANIZATIONAL CONTACT:
L. R. Hess, President TELEPHONE NUMBER:
(714) 557-1710 PRINCIPAL PRODUCT: Automatic Dampers NUCLEAR INDUSTRY ACTIVITY: Approximately 75% of the production effort is devoted to nuclear.
C ASSIGNED INSPECTOR:
()fd[
10 Af2 k-2 S2 Roberds, Reattive d Components Program Date
~
H. W.
i
= ',l Section (R& CPS)
OTHER INSPECTOR (S):
APPROVED BY:
bfit6%cy?
4 -1' $ 2-
'I. Barnes, Chie f, R& CPS '
{
Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B, and 10 CFR Part 21.
B.
SCOPE:
This inspection was made as a result of the identification to Region IV of the Nuclear Regulatory Commission of potential material and design problems in automatic dampers supplied to Grand Gulf Nuclear Station by Pacific Air Products.
Areas selected for inspection included:
drawing control; control of purchased materials, manufacturing process control and final inspection.
PLANT SITE APPLICABILITY:
Dockets:
50-416, 50-417.
119
ORGANIZATION:
PACIFIC AIR PRODUCTS COMPANY SANTA ANA, CALIFORNIA REPORI IN$PECTION NO.
99900769/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Quality Assurance Operating Procedure No. 10.0, inspection did not tag a shipment of 12 ga., 14 ga., and 16 ga. galvanized sheets, observed in the controlled storage area, with the required green accept tag.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Quality Assurance Operating Procedure No. 3.0, the Chief Engineer did not assure that design documents (drawings) contained correct and adequate application of design criteria, as evidenced by the absence on Drawing Sheet No. 5445-14, Rev. 4 for Shop Order 1383, Damper Nos. 1-CBA-DP-27A and B, of required location of welds, joint configuration and weld sizes.
3.
Contrary to Criterion V of Appendix B to 10 CFR part 50 and Quality Assurance Operating Procedure No. 10.0, an in process inspector did not tag an actuator mounting bracket on Damper No. 1-CBA-DP-27A l
as nonconforming although observed by the NRC inspector to contain cracks at both ends of a bend.
C.
UNRESOLVED ITEMS:
None D.
OTHER FINDINGS OR COMMENTS:
1.
10 CFR Part 50.55(e) Construction Deficiency Report (CDR)
The problem reported to the NRC by Mississippi Power and Light Company was tne failure of an actuator to damper shaft coupling in the HVAC automatic dampers in the Diesel Generator Building at Grand Gulf Nuclear Station.
120
ORGANIZATION:
PACIFIC AIR PRODUCTS COMPANY SANTA ANA, CAllFORNIA REPORI IN5PECIIGN NO.
99900769/82-01 RESULTS:
PAGE 3 of 3 An investigation by Pacific Air Products Company (PAPCO) revealed that the actuators were misaligned with the damper shaft wnan the unit was installed which created an additional stress on the coupler.
The actuators were installed to the damper by site personnel at the request of Bechtel Power Corporation, Grand Gulf Nuclear Station PAPC0 requested the return of all couplers from Grand Gulf and performed torque test on the returned couplers at 600% overload rate with no observed failures of the couplers.
To prevent future failures as a result of misalignment, PAPC0 has ro'iesigned the coupler by the addition of an attached sleeve to provide additional reenforcement.
2.
Material Identification and Control This area 01 the inspection was performed by review of the applicable Quality Assurance requirements, associated documentation and observations of the controlled storage area Nonconformance Item B.1 was identified.
3.
Drawing Control This area of the inspection was performed by review of the applicable Quality Assurance requirements, associated documentation and four drawings issued to production at various stages of the fabrication cycle.
Nonconformance Item B.2 was identified.
4.
Manufacturing Process Control and Final Inspection This area of the inspection was performed by review of the applicable Quality Assurance requirements, associated documentation and five production order travelers at various stages of the fabrication cycle.
Nonconformance Item B.3 was identified.
121
ORGANIZATION:
THE WILLIAM POWELL COMPANY CINCINNATI, OHIO REPORI INSPECTION INSPECTION NO.
9990E157/82-01 DATE(S) 2/22-26/82 ON-SITE HOURS: 60 CORRESPONDENCE ADDRESS:
The William Powell Company ATTN:
Mr. Paul Niehaus, Vice President of Engineering and Manufacturing 2503 Spring Grove Avenue Cincinnati, Ohio 45213 ORGANIZATIONAL CONTACT:
Mr. E. E. Winterfeldt, Corporate QA Manager TELEPHONE NUMBER:
(513) 852-2967 PRINCIPAL PRODUCT: Nuclear Valves NUCLEAR INDUSTRY ACTIVITY: Fifteen contracts for ASME, Class 1, 2, and 3 valves.
' /-
w~~
'e
' 2_
ASSIGNED INSPECTOR:
R. L. Brown, Reactive & Components Program Section Date (R& CPS)
OTHER INSPECTOR (S):
L. E. Ellershaw, R& CPS E. L. Hines, R& CPS APPROVED BY:
1 J4 ~ '
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21 and 10 CFR Part 50, Appendix B.
B.
SCOPE:
This inspection was conducted as a result of: (1) allegations made regarding improper weld repairs of hardfacing on valves supplied to Cincinnati Gas & Electric Company's Wm. H. Zimmer Nuclear Power Station, Unit 1; and (2) a Construction Deficiency Report from Mississippi Power
& Light Company pertaining to valve stem protectors creating operational problems at Grand Gulf Nuclear Station, Unit 1. Additional areas inspected (SCOPE Cont. on next page)
PLANT SITE APPLICABILITY:
While the activities of this organization relate to numerous plant sites, this inspection was limited to the activities concerning Docket Nos. 50-416; 50-417; and 50-358.
123 i
ORGANIZATION:
THE WILLIAM POWELL COMPANY CINCINNATI, OHIO REPORT If6FtL610N NO.
99900057/82-01 RESULTS:
PAGE 2 of 6 SCOPE (Cont.)
included nondestructive examination personnel qualifications, weld material control, manufacturing process control, nonconformances and corrective actions, change control and status of previous inspection findings.
A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, and QA Manual Article 5, paragraph 5.5.1.4, controlled welding materials were stored in holding ovens in the Bonded Storage areas in which they were not separated nor was identity maintained; e.g.
a.
Holding oven number 2027, bin 4, identified as containing 5/32" type 8018 electrodes, also contained 5/32" type 7018 electrodes.
b.
Holding oven number 5850 was identified as containing various sizes of type 7018 electrodes, stored in their respective bins; however, numerous unidentified electrodes were also found in this oven.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, and QA Manual, Article 2, paragraph 2.14.9.2, a number of inspection steps on certain rework routers had not been stamped in the appropriate column to signify acceptance.
C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
(0 pen) 10 CFR Part 21 Report (79-02):
1.
Mississippi Power and Light Company Grand Gulf, Unit 2.
The structural configuration changes for the 36 Unit 2 valves had not been started.
The William Powell Company (WPC) management stated that WPC is negotiating with the utility company relative to the responsibility for the modifications.
2.
Cincinnati Gas and Electric Company Zimmer Project.
WPC has completed the seismic analysis of the Zimmer Valves, using the latest approved design evaluation methods (computer program, etc).
The results of 124
f ORGANI7ATION:
THE WILLIAM POWELL COMPANY CINCINNATI, OHIO RLPORI INdPECi10N NO.
99900057/82-01 RESULTS:
PAGE 3 of 6 these analyses have been submitted to Sargent & Lundy Engineers (S&L)
(plant AE).
There has been some communcations between WPC and S&L relative to this subject, but a final disposition has not been deter-i mined.
E.
OTHER FINDINGS OR COMMENTS:
1.
Change Control:
Review of one customer change request, one change to WPC order, order addenda assignment record, controlled material bill of change, bill of change distribution sheet, addendum to seismic report, three drawings, list of material, submittal letter to customer for field conversion, and document recall requests verified conformance with the program that describes the method of entering the customer documents into the system and control of the generated documents through the plant operations.
2.
Control of Nonconformances and Corrective Action:
Review of eight trouble analysis reports, eight production rework routers; three audit reports, three corrective action requests and one material review board report verified that discrepancies are identified, corrective action taken, activities are documented and reported to management as required by the quality assurance manual.
3.
Manufacturing Process Control:
Shop activities are controlled by routers.
Review of three routing cards, process control sheets for three shop orders and nine routers verified that the pertinent manu-facturing and inspection steps were identified, the assigned witness /
hold points were appropriately signed and the completed router was signed by quality personnel signifying acceptance, however, one nonconformance was identified (see paragraph B.2).
4.
General:
WPC management stated that effective March 1, 1982, the commercial nuclear valves assigned to Plant 1 (Docket 99900269) will be transferred to Plant No. 2.
They also stated that Plant 1 will let the ASME Certificate N-1337 expire on March 1, 1982.
l They further stated that in the future Plant 1 may do subcontract work, in which case they will be treated as any other vendor on the approved vendor list.
1 5.
Follow up of Allegations:
On September 25, 1981, an allegation was made by a former employee of the William Powell Company (WPC) to the NRC's Region III inspector at the Wm. H. Zimmer Nuclear Power Station, pertaining to improper repair welding performed on the hardfacing of valve seats and discs by WPC during 1977-1978, on valves supplied to the Zimmer and Enrico Fermi Atomic Power Plant sites.
The 125 l
ORGANIZATION:
THE WILLIAM POWELL COMPANY CINCINNATI, OHIO REPOR.
IN5PECTION NO.,
99900057/82-01 RESULTS:
PAGE 4 of 6 alleger provided specific control numbers of the valves for which he was concerned; e.g., 71C, 1021C, 1028C, and 1048C.
The alleger further stated that all four valves would bear his identification number, as he had performed certain operations.
During this inspection, the NRC inspector reviewed the quality assurance documentation of the four valves in question.
a.
Valve 71C:
This valve was never manufactured.
The number, 71C, was assigned to a 4", 150 lb. valve destined for Arkansas Power
& Light Company.
A shop order was initiated on August 14, 1971, and subsequently cancelled on October 14, 1971.
No manufacturing operat'.ons were ever performed.
b.
Valve 1021C:
This valve was supplied to Mississippi Power & Light Company (MP&L).
The valve is a 4", 150 lb. weld end gate valve with a forged body.
The route sheet was initiated on June 30, 1976, and the valve was completed and sent to the bonded store room on February 14, 1977.
There was no hardf acing operation per-formed on the disc which is SA 351, stainless steel, Grade CA 15.
There was hardfacing of the carbon steel, SA-105, seat rings, with Ste11ite No. 6.
This operation was performed on February 2, 1977, with subsequent magnetic particle examination (MT) per-I formed on February 11, 1977.
The MT was a mandatory hold point which was witnessed by the customer.
c.
Valve 1028C:
This valve was supplied to MP&L and is a 3",
150 lb. weld end gate valve with a forged stainless steel body.
The route sheet was initiated or. February 7,1977.
The seat rings were assembled and welded on August 11, 1977 with subse-quent liquid penetrant examination (PT) on August 19, 1977.
Final inspection was performed and the valve body was sent to bonded stores on August 26, 1977.
The disc is a stainless steel CF 8 casting which was hardfaced on August 17, 1977, with subsequent PT performed on September 28, 1977.
Final inspection took place on October 14, 1977, and the disc was sent to bonded stores.
d.
Valve 1048C:
This valve was supplied to MP&L and is a 6", 150 lb.
swing check valve, with a cast carbon steel body, grade WCB, and a forged carbon steel disc, SA-105.
The route sheet for the body was initiated on March 25, 1976, with all operations completed on May 18, 1978.
The route sheet for the disc was initiated on March 26, 1976, with final inspection on December 23, 1976.
Hard-facing of the disc was performed on November 30, 1976, PT on December 20, 1976, and final inspection on December 23, 1976.
126 l
ORGANIZATION:
THE WILLIAM POWELL COMPANY CINCINNATI, OHIO MLrVKi ihbFLLiAUH i
NO.
99900057/82-01 RESULTS:
PAGE 5 of 6 It should be noted that the ASME Code,Section III, NB, NC, and l
ND 2121(b) exempts valve seats f rom code jurisdiction.
1 All welding procedure specifications (WPS) and their procedure i
qualification records (PQR), designated as being used during l
hardfacing operations, were reviewed by the NRC inspector.
In l
addition, the qualifications of the welders and the certified materiai test reports of the welding materials used, and the postweld heat treat charts were reviewed.
The "Hardfacing Opera-tion Record & Inspection Reports" (separate from the route sheets) were reviewed and compared with data on the route sheets for consistency related to dates, welders, and inspectors.
The nondestructive examination reports were also reviewed.
There were no welding operations performed (nuclear) during this inspection, therefore, current, in process activities could not be observed.
The allegation was not substantiated, based on the documentation review of the four valves in question.
6.
Construction Deficiency Report (CDR)
Mississippi Power & Light Company (MP&L) notified NRC's Region II office on March 14, 1980, of a CDR pertaining to the potential for valve stem protectors to cause valve misoperation due to excessive threading of the protectors.
This would allow the protector to be screwed too far into the valve operator, thus acting as a jam nut against the stem lock nut.
This problem was associated with WPC valves.
WPC was not formally notified by MP&L or Bechtel, the architect engineer and constructor.
An internal, Bechtel memo dated March 14, 1980, dealing with stem protectors on motor operated valves, was provided to WPC.
In addition to the memo, Bechtel Management Corrective Action Report No. 65 dated March 14, 1980, and Startup Field Report No.
1-M-385 dated March 3, 1980, were provided to WPC.
WPC, because of this information, initiated a documented inspection program to measure the clearances between the stem protectors and the stem lack nut on all motor and adaptor gear operator valves.
This inspection was initiated in May 1980, and continued through July 1980.
Approximately 150 valves were inspected and there was not one case where the stem protector created an interference; thus, the inspection program was terminated.
(
127
ORGANIZATION:
THE WILLIAM POWELL COMPANY CINCINNATI, OHIO REPORI IN$PLLIION NO..
99900057/82-01 RESULTS:
PAGE 6 of 6 Observation of two nuclear valves with stem protectors revealed that the stem protectors could not be forced in far enough to create an interference.
However, WPC agreed to reinstitute the documented inspection program to eliminate any possibility of this type problem.
7.
Nondestructive Examination (NDE) Personnel Qualifications The qualifications of seven NDE personnel were reviewed, including results of written examinations and eye examinations.
The review included four NDE disciplines:
magnetic particle examination; liquid penetrant examination; radiography; and ultrasonic examination.
This area appeared to be in conformance with WPC's QA Program commitments.
8.
Weld Material Control This area of the inspection was performed by:
observing weld material storage, including holding ovens; review of 10 weld material test reports associated with the electrodes identified in the holding ovens; and review of the welding material issue system.
Nonconformance B.1 was identified during this portion of the inspection.
I i
t 128
ORGANilATION:
SIEMENS-ALLIS INC.
SMALL MOTOR DIVISION LITTLE ROCK, ARKANSAS REPORT INSPECiION INSPECTION NO.
99900338/82-01 DATE(S) 2/16-18/82 ON-SITE HOURS: 16 CORRESPONDENCE ADDRESS:
Siemens-Allis, Inc.
Small Motor Division ATTN:
Mr. William E. Onyett Manager, Quality Assurance 1400 Dineen Drive Little Rock, AR 72206 ORGANIZATIONAL CONTACT:
Mr. William E. Onyctt, Manager, Quality Assurance TELEPHONE NUMBER:
(501) 897-4905 PRINCIPAL PRODUCT: Small Electric Motors, 5 HP - 75 HP NUCLEAR INDUSTRY ACTIVITY: Commercial nuclear production of small electric motors Class 1E, 5 HP to 75 HP.
At present all nuclear orders have been filled and shipped.
No commercial nuclear contracts are in house.
/
/Mvh F
YMD ASSIGNED INSPECTOR:
J. W. Sutton, Reactive & Components Program Date Section (R& CPS)
OTHER INSPECTOR (S):
P APPROVED BY:
M
/. ) m,, u-7 ' W-4 2-I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and 10 CFR Part 21.
B.
SCOPE:
Status of previous inspection findings and vendor activities.
PLANT SITE APPLICABILITY:
Not identified.
129
ORGANIZATION:
SIEMENS-ALLIS, INC.
SMALL MOTOR DIVISION LITTLE ROCK, ARKANSAS RLPORI IN3PtLi10N NO.
99900338/82-01 RESULTS:
PAGE 2 of 3 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None D.
STATUS OF PREVIOUS INSPECTION FINDINGS 1.
(Closed) Violation (80-01):
Current copies of 10 CFR Part 21 not posted.
The inspector verified that current copies of 10 CFR Part 21 have been posted on all office and factory bulletin boards.
2.
(Closed) Deviation A (80-01):
Index checklist for audit report and audit summary had not been formulated or included in revisions to forms.
The inspector verified by review of five nuclear audits, revised QA procedures and forms that the commitments made in Siemens-Allis' February 16, 1981, reply to NRC's letter of February 4, 1981, have been satisfactorily implemented.
3.
(Closed) Deviation B (80-01):
Audit reports had not been completely filled out as required by Procedure 10.01.
The inspector reviewed ten audit reports issued during 1981 for compliance to Siemens-Allis' (SA) commitment for filling out audit reports.
The reports reviewed were found to be completed as required by SA's QA procedures.
4.
(Closed) Deviation C (80-01):
Quality Assurance had not audited or applied " Hold Tags" on certain identified motcrs.
130
ORGANIZATION:
SIEMENS-ALLIS, INC.
SMALL MOTOR DIVISION LITTLE ROCK, ARKANSAS REPokl IN5Ptul1UN NO.*
99900328/82-01 RESULTS:
PAGE 3 of 3 The inspector verified by examination of available shop records, audits and work in progress that " Hold Tags" are being used to identify defective parts or tests for repair or rework.
The QA inspector is stamping those operations that he has witnessed.
The inspector examined six " Hold Tags" for conformance and all were found to have been completed as required by SA QA procedures.
Random audits have been conducted by QA personnel for compliance.
5.
(Closed) Deviation D (80-01):
Measures had not been established and documented to assure control and distribution of documents, including changes to the location or person using the documents.
The inspector reviewed current test specifications, and test station engineering copies of tests, weekly audits, and other available documentation to verify that current documents have been distributed as required by QA procedures.
The documents reviewed were found to be in order and comply with required distribution.
6.
(Closed) Deviation E (80-01):
File copy of operator instruction sheets had not been approved by production or manufacturing.
The inspector reviewed a QA audit of all master NSRE-0IS sheets for signature.
The inspector verifed that all previous sheets had been signed off as required.
Five current sheets and master sheets were reviewed and found to have been signed off as required.
7.
(Closed) Deviation F (80-01):
Aperture cards from product engineering file had not been removed or marked as required by procedure.
The inspector reviewed the aperture file to determine if procedures are being followed.
Procedure GP 15 has been revised to clarify para-graph 4.1.4 requirements.
QA has conducted audits to assure compliance to the revised requirements.
8.
(Closed) Deviation G (80-01):
Item sheets had not been signed off and the person checking the Test Data Sheets had not noted " verified" by his name.
The inspector verified by review of ten current item sheets that the signatures of all required personnel have been recorded and verified.
SA QA personnel have conducted audits of this area to assure compliance with procedure requirements.
131
ORGANIZATION:
SQUARE D COMPANY CONTROL GROUP RALEIGH, NORTH CAROLINA RLPORI INSPECTION INSPECTION NC.
99900717/82-01 DATE(S) 3/8-12/82 ON-SITE HOURS: 28 CORRESPONDENCE ADDRESS:
Square D Company Control Group A T IN:
Mr. D. J. Beck, Plant Manager P. O. Box 2/446 Raleigh, NC 27611 ORGANIZATIONAL CONTACT:
Mr. W. J. Fightmaster, QA Manager IELEPHONE NUMBER:
(919) 266-3671 PRINCIPAL PRODUCT: Class 1E starters and contractors.
NUCLEAR INDUSTRY ACTIVITY: Commercial Nuclear Production of Square D Company, Control Group, Raleigh, North Carolina, totals less than 1% of total company production.
Two nuclear contracts are presently in-house for future production.
ASSIGNED INSPECTOR: _h Wh 4 - 2.-$ 2 J MW. Sutton, Reactive & Component Programs Section Date (R& CPS)
OIHER INSPECTOR (S):
9 ' A o- - -,
m APPROVED BY:
+ - s - r.?._.
I. Barnes, Chief, R& CPS Date
~
INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
Follow up previous inspection findings, nonconformance/ corrective action, audits and review of vendor activities.
PLANT SITE APPLICABILITY:
Not identified.
133
i l
i ORGANIZATION:
SQUARE D COMPANY CONTROL.iROUP RALEIGH, NORTH CAROLINA REPUHi INaFtLituh NO.
99900717/82-01 RESULTS:
PAGE 2 of 4 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
None C.
UNRESOLVED ITEMS:
None 1
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Deviation A.1&2 (Report 80-01):
Setup and Run Sample Inspection Sheet not completed as required.
Run checks not conducted or required.
The inspector reviewed the corrective action taken by Square D Company to prevent recurrence.
Thirteen shop Setup She( s completed since issuance of an instruction letter dated Jure 25, 1980, were reviewed for compliance.
As a result of the review, the inspector determined that inspectors are complying with the inspection procedure and signing the Setup Sheets as required by QA instructions and a Square D letter of June 25, 1980.
2.
(Closed) Deviation B.1&2 (Report 80-01):
Name of issuing engineer not shown in the deviation stamp for extension of a drawing deviation.
The inspector reviewed the corrective action and preventive measures taken by Square D Company to prevent recurrence of this item.
The inspector checked five drawings with current deviations to deter-mine that the engineer is signing the deviated drawing.
In addition, QA audits of the deviation files have been performed for compliance to QA procedures / instructions.
As a result of the review of the above documents, the inspector has determined that drawings are being controlled as required by QA procedures.
3.
(Closed) Deviation C (Report 80-01):
Weekly Delinquent Reports did not include the number of problem reports initiated and resolved during the month, nor the number of problem reports pending.
134
ORGANIZATION:
SQUARE D COMPANY CONTROL GROUP RALEIGH, NORTH CAROLINA RLPORI 1h5PtGi10N NO..
99900717/82-01 RESULTS:
PAGE 3 of 4 The inspector reviewed the revised Standard Practice Instruction (SPI) 655.1-M29 dated June 27, 1980, for revised changes.
In addition, the inspector reviewed six current delinquent Materials Reports to deter-mine if the revised instructions are being followed.
The inspector has determined that corrective action and preventive measures instructions are being implemented as required in the SPI.
4.
(Closed) Deviation D.1,2,3,4 (Report 80-01):
Failure to maintain identifying numbers, cards, recalibration schedules, and calibration stickers on equipment in the controlled Square D calibration system.
The inspector reviewed the corrective action / preventive measures taken by Square D Company to correct this item.
The following areas were checked for conformance.
Letter of June 23, 1980, to all tool and gage inspectors to maintain calibration control.
Review of the cardex system for control of calibration.
The inspector also exa-mined 6 gage control cards for compliance, and examined 10 tools, gages, and bench boxes for calibration stickers and calibration dates.
The inspector also reviewed all QA audits conducted for compliance to preventive action commitments.
As a result of the review of documentation, the inspector determined that activities relating to control of calibration of equipment now conform to QA manual and instruction requirements.
5.
(Closed) Deviation E (Report 80-01):
Basic Device Gaging Fixture and accompanying gage blocks not calibrated to provide assurance that they were of the required accuracy and in acceptable operating condition.
The inspector reviewed the corrective action and preventive measures taken by Square D Company to determine if any other nonidentified gages were in the final inspection / acceptance areas.
The inspector conducted inspection of these areas and found that acceptance gages, tools, etc., had been calibrated and were being controlled.
The QA instructions were being complied with, as required.
6.
(Closed) Unresolved Item C.3.b (Report 80-01):
Manufacturing process control.
There is no requirement to maintain records of routing checks to ML-STD-1050.
The inspector reviewed five Routing Check Sheets to determine if records of routing checko are being completed as required by instructions.
As a result of this review, the inspector determined that QA instruc-tions ace being followed as required.
135
ORGANIZATION:
SQUARE D COMPANY CONTROL GROUP RALEIGH, NORTH CAROLINA REPORT INSPECTICN NO.
99900717/82-01 RESULTS:
PAGE 4 of 4 7.
(Closed) Unresolved Item D.3.b:
Procedure to control salvage activities not available.
The inspector reviewed minutes of weekly meetings held witti the QA Manager, Plant Supervisor, General Supervisor of Areas, and General Supervisor Inspection, conducted to determine and follow up on identified salvageable materials that would be returned to stock.
Management has decided that a procedure to control this item is not necessary.
After review and discussion with Square 0 management personnel, the inspector concurred that the control of salvageable materials was being handled in an acceptable manner.
D.
OTHER FINDINGS:
1.
Nonconformance/ Corrective Action - The inspector reviewed Squarc D's QA Manual instructions 653-M-165, R655-E-132b-R655.1 and R655-E-126b covering control and disposition of no.iconforming items. The inspector reviewed 6 nonconformances,10 Weekly Deficiencies Reports and 11 Quality Problem Reports and associated documentation for con-formance to QA/QC instructions.
The corrective action taken to prevent recurrence was examined for implementation.
Segregation areas designated for storage of nonconforming items were inspected.
Relative to the documents examined, the inspector determined that activities relating to the control and disposition of nonconforming items complied with Square D's QA/QC program requirements.
2.
Audits - The inspector reviewed Square D's QA Manual, SPI-E-136b, Audits for Compliance to 10 CFR Part 50 Requirements.
Twenty com-pleted audits, corporate group audits, interoffice memos of QA activities and corrective action taken as a result of the audits conducted were reviewed for compliance to QA instructions.
As a result of the review, the inspector determined that the audit requirements, both corporate and plant, were being followed as required by QA instructions.
I 136 i
ORGANIZATION:
STONE AND WEBSTER ENGINEERING CORPORATION BOSTON, MASSACHUSETTS ktruki iNSPLLl10N INSPLCl10N l
NO.
's9900509/82-01 DATE(S) 3/8-12/82 & 3/22-26/8? ON-SITE HOURS: 111 l
l CORRESPONDENCE ADDRESS:
Stone and Webster Engineering Corporation ATIN:
Mr. R. B. Keliy l
Vice President, Quality Assurance P. O. Box 2325 Boston, MA 02107 l
ORGANIZATIONAL CONTACT:
Mr. F. B. Baldwin, Assistant QA Manager TELLPHONE NUMBER:
(617) 973-6566 PRINCIPAL PRODUCT:
Architect Engineering Services NUCLEAR INDUSTRY ACTIVITY: Major projects include Beaver Valley 2, River Bend 1 and 2, Shoreham, Nine Mile Point 2, Millstone 3, North Anna 3, New Haven 1 and 2, Montague 1 and 2, and Jamesport 1 and 2.
In addition, there are 44 modification / repair / service contracts.
The aforementioned contracts cover work performed in the Boston, Cherry Hill, New York, and Denver offices.
(,[h'l ASSIGNED INSPECTOR:
tt c
D. D. I mLerlain, Rea/ tor Systems Section (RSS)
[Jatd OTHER INSPECTOR (S): A. L. Smith, Equipment Qualification Section J. I. Tapia, Engineering Section
[
(
[-!'/
'L
\\
APPROVED BY:
C. Jrp fe, ChieT RSS (f a t6 '
INSPECI10N BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and Topical Report SWSQAP 1-74A.
B.
SCOPE:
Status of previous inspection findings (Boston and Cherry Hill offices), design process management (follow up on concern identified in 81-03 report at Boston office), high energy line ruptures in fluid systems outside containment (Boston office), safe shutdown following pipe rupture outside containment (Cherry Hill office), and an (cont. on next page)
PLANl SITE APPLICABILITY:
Docket Numbers 50-404, 50-334, 50-412, 50-458, 50-459, 50-410, and 50-423.
137
ORGANIZATION:
STONE lND WEBSTER ENGINEERING CORPORATION BOSTON, MASSSACHUSETTS REPORI INSPLCi10N NO.
99900509/82-01 RESULTS:
PAGE 2 ef 7 SCOPE:
(Cont.)
inspection condet.ted as a result of the following 10 CFR Part 50.55(e) reports:
(1) 10 CFR Part 50.55(e) Report (North Anna 3) - Pressurizer safety i
salve inlet piping design with large pressure drop could lead to valve chatter (Boston of fice); (2) 10 CFR Part 50.55(e) Report (River Bend 1 and 2) - Heat loads provided by General Electric to Stone and Webster (S&W) for sizing HVAC in the High Pressure Coolant System (HPCS) diesel generator room were about one-third of the expected valve (Cherry Hill office); and (3) 10 CFR Part 50.55(e) Report (Beaver Valley 2) - Structural calculations for control room ex. tension contained errors (Boston of fice).
A.
V10 TAT 10NS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and Engi-neering Departinent Memorandum No. 82-3, all required experience records for designers in the electrical and the engineering l
mechanics division (Boston office) were not prepared or updated by March 1, 1982, as committed to the NRC.
2.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and Engi-neering Assurance Procedure 5.3, an assumption made in calculation No. 12179-US(B)-221 was not confirmed and the assumption was not identified as an assumption that required confirmation at a later date (Millstone 3 project).
3.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and Engi-neering Assurance Procedure 5.3, HVAC calculation PB-196 for control building cooling load contained input information that was not contirmed and "Yes" was not checked to indicate confirmation was required at a later date.
C.
UNRESOLVE,0 ilEMS:
Nonc 138
5I' Ni ANU nEtbTik I NG!hEERING E'PGRAT ION OEANI/All%
J E05 ION. L'AE ACHUM IT 5 HLPUPI INSPLLiibh NO.
'h9Jdb09/92-01 RESULT 5:
PAGE 3 of 7 i
D.
STATUS GF PREVIOUS INSPECIIGN FINDING 5:
1.
(Clcsed) Unresolved Item (81-03):
it was not always apparent th it the ANSI N.43.?.12 requitements and the S&W procedural requirements for holding a postaudit conference and identity-inq porcons attending the postaudit conference are hoing followed.
The NRC inspector examined a random sample of nine audits and seritied that ali reports contained documentation of both a postaudit conteronce and a list of attendees.
It is concluded from this review that S&W is meeting these requirements of ANSI N45.2.12.
2Property "ANSI code" (as page type) with input value "ANSI N45.2.12.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
(0 pen) Nanconf ormance (81-04):
Experience Records were not being updated on a yearly basis as required.
S&W management had committed to having all required Experience Records updated by March 1, 1982.
The NRC inspector examineJ a total of 46 experience records for both engineers and designers in the Boston and Cherry Hill offices.
It was determined that the electrical and engineering mechanics divisions in the Boston office had not completed the update of all experience records for designers as committed.
This item will remain open at the Boston office and a nonconformance (B.1' above) was issued for the failure to meet a commitment to the NRC.
E.
OTHER FINDINGS OR COMMENTS:
1.
Design Process Management (Follow up on concern identified in 81-03 report) - During the 81-03 NRC inspection, a discrepancy was identi-fIed on the Millstone 3 project relating to an instrument set point calculation for the component cooling water system.
The calculation was a " trial" calculation for a new instrument set point calculation system that was being developed.
A procedure for this system was issued on March 8, 1982, and a small number of calculations have been completed.
From the review of this area, it was determined that the new set point calculation system is being implemented and should provide good control in the future.
2.
High Energy Line Ruptures in Fluid Systems Outside Containment THoston office) - The applicable SAR sections, procedures, and criteria for high energy line (HEL) rupture analysis were examined 139
ORGANIZATION:
STONE AND WEBSTER ENGINEERING CORPORATION BOSTON, MASSACHUSETTS REPORI IN5PECIl0N NO.
99900509/82-01 RESULTS:
PAGE 4 of 7 to determine program requirements.
The following design documents for the Millstone 3 project were examined to verity implementation of program requirements:
eight design drawings, nine analysis cal-culations, and one specification.
From this area of inspection, it was determined that the guidance provided in Branch Technical Positions MEB 3-1 and APCSB 3-1 are being followed for selecting pipe ruuture locations, rupture sizes and geometry.
A noncon-formance was identified (B.2 above) regarding confirmation of assumptions in a pressure / temperature analysis calculation.
The assumption stated that all areas sealed with weather-stripped doors are not exposed to high energy line breaks from adjacent areas.
This assumption was not identified as an assumption that must be confirmed and there was no confirmation made to assure that the doors specified would withstand the pressures generated in adjacent compartments.
Due to the incomplete status of the HEL analysis for the Millstone 3 project by S&W, our inspection effort in this area could not be completed during this inspection.
3.
Safe Shutdown Following Pipe Rupture Outside Containment (Cherry Hill of fice) - The review of HEL documentation on the River Bend project indicated that the break criteria contained in Branch Technical Positions APCSB 3-1 and MEB 3-1 were used as the basis for postu-lating pipe rupture conditions.
The attempted review of documenta-tion containing the plant transient analyses that support the posi-tion that the plant can be safely shutdown in the event of any pipe rupture indicated that the SAR Chapter 15 accident analyses are viewed as the bounding analyses as long as the safe shutdown systems are protected from the effects of pipe rupture.
This area will be inspected further during subsequent inspections.
4.
High Pressure Coolant System (HPCS) Diesel Generator Room Heat Loads - This area of inspection resulted from a 10 CFR Part 50.55(e) report (River Bend 1 and 2) which stated that the heat loads provided by General Electric (GE) to S&W for sizing the heating, ventilation, and air conditioning (HVAC) in the HPCS diesel generator room were about one third of the expected value.
S&W had questioned the heat load values provided by GE because they were significantly lower, per kilowatt, than the values being used for the main plant diesels.
GE is presently evaluating this item and S&W has not officially been l
130 l
I ORGANIZATION:
STONE AND WEBSTER ENGINEERING CORPORATION BOSTON, MASSACHUSETTS REPORI IN5PLCIION NO.
99900509/82-01 RESULTS:
PAGE 5 of 7 provided with new values.
However, S&W evaluation has convinced them that the values are low and S&W has decided to revise the HVAC design to provide ventilation for the HPCS diesel generator room that equals the main diesel generator room capacity.
S&W also had responsibility for design of HVAC for the Nine Mile Point (NMP) 2 project HPCS diesel generator room but the original design called for HVAC capacity equal to the main diesel generator room, therefore the design is adequate for NMP-2.
An examination of the HPCS diesel generator room HVAC calculations revealed that S&W program require-ments were implemented.
The HVAC calculation for the Control Building Cooling Load (PB-196) was also examined to determine the source of heat load data and the required confirmation of input data used.
A nonconformance was identified (B.3 above) regarding confirmation of input data.
The inputs noted that require confir-mation were GE supplied information that was identified as preli-minary and heat load information for computer equipment.
The specific nonconformance was corrected during this inspection by revising the calculation to indicate confirmation was required.
It was also noted that S&W had performed an internal audit during March 1982, which identified similar concerns with HVAC calcu-lations. However, S&W had not determined corrective actions and preventive measures for the internal audit finding, therefore, a response to the NRC nonconformance will be required.
The HPCS diesel generator heat load item will remain open pending the completion of the GE evaluation and possible generic implications at other Architect Engineers.
S&W actions for affected projects is adequate and no further review will be required at S&W.
5.
Potential CDR Pressurizer Safety Valve Problem at North Anna, Unit 3 -
The Electric Power Research Institute (EPRI) is currently performing a study of pressurizer safety relief valves and piping systems.
The original study's intent was to quality individual safety relief valves by testing various sizes and piping arrangements for safety relief valves representative of all reactors in operation or under construction.
The original intent has had to be revised since representative safety relief valves ha been found to interact with piping configurations (in particular tne characteristics of piping upstream of the valve).
EPRI contacted utilities to determine plant 141
ORGANIZATION:
STONE AND WEBSTER ENGINEERING CORPORATION BOSTON, MASSACHUSETTS REPORI IN5Phtl10N NO.
99900509/82-01 RESULT 5:
l PAGE 6 of 7 specific safety relief valve inlet pipe configurations and performed a preliminary analysis using computer modeling to predict safety relief valve performance for the various configurations.
The pre-liminary calculations performed under this study indicate that valve chatter (as a result of the pressure drop in the inlet piping) is expected on the North Anna, Unit 3, system.
Virginia Electric and Power Company (VEPCO) had their A/E (Stone & Webster) review the EPRI calculations.
The review performed by S&W indicated no inconsistencies in the EPRI calculation, hence, VEPC0 notified Region II of a potential deficiency under the provisions of 10 CFR 50.55(e) and has instructed S&W to revise the design of the inlet piping to permit satisfactory valve operation.
S&W has entered this problem into their problem reporting system and VEPC0 was informed that (a) the stress analysis of the pressurizer safety relief valve inlet piping has not been performed, (b) when performed the detailed stress analysis will consider the response of the piping system due to a valve opening transient condition of the type addressed by EPRI, and (c) if a detailed analysis indicates a degradation et system performance, the appro-priate system component (s) will be redesigned to rectify the potential problem.
This item will remain open since this was a result of preliminary analysis and S&W prefers to have, in hand, the final EPRI report prior to initiating any redesign.
In addition, a decision concern-i l
ing increasing the height of the pressurizer is pending and any l
increase in pressurizer height would automatically result in a redesign of the inlet piping configuration.
We will continue to monitor this potential problem for its generic aspects, since the preliminary EPRI findings indicated that several more plants may also experience probable safety relief valve problems, particular for the short valve opening times.
6.
Structural Design Deficiencies at Beaver Valley, Units 1 and 2 -
On August 31, 1981, Duquesne Light Company submitted a 10 CFR Part 50.55(e) report concerning structural design l
deficienies identified during the review of the design calcula-I tions for the Unit 2 control room extension.
As a result of this discovery, S&W conducted a reverification of all calculations
{
l 142
l ORGANIZATION:
STONE AND WEBSTER ENGINEERING CORPORATION BOSTON, MASSACHUSETTS REPORI INSPECIION NO.-
99900509/82-01 RESULTS:
PAGE 7 of 7 performed by the group of engineers involved in the original design.
The reverification program disclosed a discrepancy in the design of the Auxiliary Building floor slab supporting the boron recovery test tanks.
The NRC inspector performed an evaluation of the generic implica-tions of all work performed by the group of engineers responsible for the design deficiencies.
The evaluation included a review of the S&W problem report system documentation of the reverification of work.
It was determined that the group of engineers was only involved in a limited scope of work in the Beaver Valley Project.
The original calculations were performed in 1974 and 1975 in the New York office of S&W.
The discrepancies involved failure to include loads due to flooding, failure to check drawings against calculation requireiaents for accuracy, and a failure to justify an engineering assumption certaining to load distribution.
The S&W Engineering Assurance Division, in January of 1977, revised Engineering Assurance Procedure No. 5.3, " Control of Computerized and fianual Calculations," to include the requirement in NRC Regula-tory Guide 1.64, " Quality Assurance Requirements for the design of Nuclear Power Plants" for an independent review of design calcula-tions.
The required independent design verification represents a more formalized program than that existing at the time of the discrepancies.
By review of ongoing design activities during this inspection, the NRC inspector found that preventive measures now exist which are intended to preclude similar design deficiencies from being undetected.
I 143
ORGANIZATION:
TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS RLPURI IN$PECTION INSPECTION NO.
99900513/82-01 DATE(S) 3/29-4/2/82 ON-SITE HOURS: 79 CORRESPONDENCE ADDRESS:
Teledyne Engineering Services ATTN:
Mr. F. C. Bailey President 130 Second Avenue Waltham, MA 02254 ORGANIZATIONAL CONTACT:
Mr. C. G. Sprangers, Manager, Quality Assurance TELEPHONE NUMBER:
(617) 890-3350 PRINCIPAL PRODUCT: Engineering and Consulting Services.
NUCLEAR INDUSTRY ACTIVITY: Approximately 90% of the staff of the Waltham, Massachucetts, facility and 30% of the Hayward, California, facility are involved in nuclear activities.
Major projects include work on Turkey Point Units 3 and 4, V. C. Summer Unit 1, Fermi Unit 2, Limerick Unit 1, and Diablo Canyon Unit 1.
1 ASSIGNED INSPECTOR:
h-(
2 O
b D. F.
Fp$yTRebc' tor Syst/ds Section (RSS)
Date OTHER INSPECTOR (S): A.
L. Smith, Equipment Qualification Section APPROVED BY:
[
Q b
C. J.(Haje', Chief, RSS Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
This inspection was made as a result of concerns expressed to NRC pertaining to possible deficiencies in the analysis for structural integrity of the torus catwalk and piping of the Fitzpatrick and Vermont Yankee Nuclear Power Plants; to cvaluate the in place quality assurance program; and conduct an initial management meeting.
PLANT SITE APPLICABILITY:
J. A. Fitzpatrick, Docket 50-333; Millstone Unit 1, Docket 50-245; Nine Mile Point Unit 3, Docket 50-220; Pilgrim Unit 1. Docket 50-293; Vermont Yankee, Docket 50-271; Diabic "'nyon, Docket 50-275; and Enrico Fermi 2, Docket 50-341.
145
ORGANIZATION:
TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS REPORT INSPECIl0N NO.
99900513/82-01 RESULTS:
PAGE 2 of 6 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to the Criterion V of Appendix B to 10 CFR Part 50 and the Diablo Canyon Project Program Plan, QA records were not stored in a single record storage facility which meets the imposed require-ments of ANSI /ASME NQA-1-1979, nor were the duplicate records stored in separate locations.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and section 3.0 of the QA Manual (QAM), design activities related to Mark I contain-ment torus hydrodynamic analyses and design modifications for the J. A. Fitzpatrick, Millstone, Nine Mile Point, Pilgrim and Vermont Yankee Nuclear Power Plants were not being accomplished in accordance with the QAM.
Specific examples of this nonconformance are:
(a) Hydrodynamic analyses did not include sufficient referencing of source data, principles and assumptions to permit ready traceability as required by section 3.6.1 of the QAM.
Further, the checker of hydrodynamic analyses did not perform the duties prescribed in section 3.6.2 of the QAM as required by section 3.6.1 of the QAM.
(b) Calculations exhibiting the signatures of the originator, checker and the design verifier were not treated with the status of a QA record as required by section 3.6.3 of the QAM.
3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and the Project Quality Assurance Program (PQAP) for the J. A. Fitzpatrick, Millstone, and Vermont Yankee Nuclear Power Plants, activities affecting quality regarding Design / Analysis Control, Project Personnel Assignment, and Project General (Engineering) Control were not accomplished in accordance with prescribed procedures in that the required procedures were either not imposed or not being implemented on the above projects.
146
i 1
ORGANIZATION:
TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS REPORI INSPLC110H NO.
99900513/82-01 RESULTS:
PAGE 3 of 6 4.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and the PQAP's for the Fermi 2, Millstone and Vermont Yankee projects, audits were not accomplished within the specified intervals, l
nor were they waived in accordance with the prescribed conditions and requirements.
5.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 17 of the QAM, PQAF's for eight projects contained a requirement to retain audit records for a period of 1 year, rather than for 6 years as required by the QAM.
C.
UNRESOLVED ITEMS:
None D.
OTHER FINDINGS OR COMMENTS:
1.
Management Meeting - The purpose, scope, and nature of the Licensee, Contractor and Vendor Inspection Program were reviewed with the President of Teledyne Engineering Services (TES) and the Quality Assurance Manager.
The concerns expressed to NRC with respect to possible errors and deficiencies in the hydrodynamics load analyses of General Electric (GE) Mark I containment torus components, and in the subsequent design of related component supports, were outlined.
The format, :ontent, dissemination and publication of NRC inspection reports, and TES responses thereto, were discussed in detail.
2.
Possible Analytical Deficiencies - An inspection was conducted to determine the validity and safety significant of concerns expressed to NRC that the use of unqualified individuals, inaccurate analy-tical models, erroneous input to structural calculations and excess management schedular pressure could result in possible deficiencies f
in the analyses for structural integrity, and in the subsequent design of internal and external component supports for the torus and its associated catwalks and piping of the Fitzpatrick and Vermont Yankee Nuclear Power Plants.
Review of TES records, and interviews with cognizant personnel, indi-cate that TES was contracted by the owners of the Fitzpatrick, Millstone Unit 1, Nine Mile Point Unit 1, Pilgrim Unit 1 and Vermont Yankee Nuclear Power Stations, in 1975-1976, to provide consulting services and to perform certain analyses and design 147
ORGANIZATION:
TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS REPORT INSPECTION j
NO.
99900513/82-01 RESULTS:
PAGE 4 of 6 modifications pertaining to the torus of their Mark I containments.
Some design modifications have already been supplied to and incorporated by the affected plants; however, the total contract effort is not scheduled for completion until late August 1982.
Review of employment, training, and qualification records, and interviews with selected engineering, management, and " contract" personnel, did not confirm the concerns that unqualified individuals were employed to do this work nor that TES management subjected the analysts and designers to schedular pressures such that the quality of their work was compromised.
However, the NRC inspector noted that TES did not have documented instructions or procedures to assure that the technical qualifi-cations (education, training, and related experience) claimed by newly hired permanent or contract employees conducting safety-related activities, was verified, stored with the status of a QA record, or that appropriate corrective action was taken when anomalies or inconsistencies were uncovered.
TES management stated that appropriate procedures would be developed and implemented prior to June 30, 1982.
This item will be followed during future inspections.
Detailed examination of the structural analyses, the resulting design modification drawings and their supporting stress calcula-tions, and interviews with cognizant personnel, indicate that the allegations relative to use of inaccurate analytical models and erroneous input data to structural calculations were based on factual observations made during a 4-month period in early 1981.
With respect to the use of inaccurate analytical models, the NRC inspector determined that the analysts:
(1) used the structurally similar Millstone torus drawing to obtain needed dimensions as inputs to Pilgrim hydrodynamic scoping and analyses since all requisite Pilgrim data was not available at the time; (2) changed the analysis techniques from a dynamic analysis to an allegedly equivalent static analysis without documented justification; 148 l
l ORGANIZATION:
TELEDYNE ENGINEERING SERVICES WAtTHAM, MASSACHUSETTS HLPORI INSPECTION i
NO.
99900513/82-01 RESULTS:
PAGE 5 of 6 (3) used preliminary and undocumented hydraulic forces as inputs to the calculations; (4) did not document the sources of design inputs nor assumptions in the calculation packages.
TES management stated that:
(1) these deficiencies are unique to work related to the identified plants and, therefore, are not generic in nature; and (2) the proper Pilgrim dimensional data has already been incorporated into Pilgrim calculations.
In response to nonccnformance B.2 above, TES committed to the following corrective actions:
(1) the engineering justification for using a static analysis method would be documented; (2) all affected calculations would be redone using the latest GE published hydraulic forces and applicable analyses methodology; (3) sources of design inputs and assumptions would be documented in the calculation packages; (4) all resulting design modification drawings and supporting stress calculations, including those previously transmitted to the affected plants would be reviewed and revised as needed.
The above actions will be completed by August 31, 1982.
One other nonconformance was identified in this area of the inspec-tion.
(See item B.3 above.) This area will be further examined during a future inspection.
Furthermore, the NRC inspector noted that Revision 0 of Technical Engineering Procedures TEP-2-004, TEP-3-002, TEP-3-003, TEP-6-008, and TEP-7-004 were not reviewed and approved by the QA Manager as required by section 3.2.1 of the QAM as evidenced by the lack of signature of the QA Manager on the subject procedures.
Since later revisions of these, and other prosedures, were reviewed in accordance with the QAM, and since the QA Manager agreed to l
document his review and approval of the subject procedures, no nonconformance was identified.
This item will be followed in a future inspection.
3.
QA Program Evaluation - The TES Corporate Policy Manual, Quality l
Assurance Manual, unique PQAP's, and the related detailed imple-menting procedures governing the areas of QA Program, organization, l
140
ORGANIZATION:
TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS REPORT INSPECIl0N
)
NO.
99900513/82-01 RESULTS:
PAGE 6 Of 6
}
cngineering control, procurement control, QA records, and audits were reviewed to determine that they were consistent with the quality and technical requirements that have been imposed on TES.
The documentation of completed work in these areas, con-sisting of training records for 14 individuals,1 manage-ment evaluation of the QA program, 8 internal QA audit files, 2 external QA audit files, qualification records for 2 audit team leaders, 4 purchase orders for TES services, 5 TES purchase orders to subcontractors, 3 drawing files, and applicable QA records, were examined to verify program imple-mentation.
Three nonconformances were identified in this area of the inspection.
(See items B.1, B.4, and B.5 above.)
150 l
ORGANIZATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND, CALIFORNIA REPORT
'N5PECIION INSPECTION NO.
99900334/82-01 DATE(S) 1/25-29/82 ON-SITE HOURS: 56 CORRESPONDENCE ADDRESS:
Transamerica Delaval, Inc.
Engine and Compressor Division ATTN:
Mr. C. Mathews, General Manager 550 85th Avenue Oakland, CA 94621 ORGANIZATIONAL CONTACT:
Mr. R. E. Boyer, Manager, Quality Assurance TELEPHONE NUMBER:
(415) 577-7422 PRINCIPAL PRODUCT: Standby Diesel Generators NUCLEAR INDUSTRY ACTIVITY: Commercial Nuclear Production of Transamerica Delaval, Inc., Oakland, California, of Standby Diesel Generators.
3000-13,000 HP totals 12% of total company production.
Two contracts are presently in-house for current and future production.
ASSIGNED INSPECTOR:
5 J. W. Sutton, Reactive & Components Program Section Date (R& CPS)
OTHER INSPECTOR (S): W. E. Foster, R& CPS 1
APPROVED BY:
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
l A.
BASES:
10 CFR Part 50, Appendix B, and 10 CFR Part 21.
j B.
SCOPE:
This inspection was made to review the status of previous inspection findings and also as a result of the issuance of 10 CFR 50.55(e) reports to the NRC pertaining to:
(1) potentially defective valve springs in i
diesel generators that have been fur ished to the following nuclear generating stations, (a) Bellefonte, Unit Nos. 1 and 2, (b) Hartsville A and B, Unit No. 2, and (c) Phipps Bend, Unit Nos. 1 and 2; and (2) a cracked (Cont. on next page)
PLANT SITE APPLICABILITY:
Potentially Defective Valve Springs:
50-519, 50-521; 50-553, 50-554; 50-458; 50-400, 50-401, 50-402, 50-403.*
Cracked Stud and Spherical Washers: 50-322; 50-416, 50-417; 50-206, 50-361, 50-362.*
Inadequate Starting Air Check Valve:
(Cont. on next page) 151 i
ORGANIZAi!ON:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND, CALIFORNIA REPORI IH3Fttiiun
(
NO.
99900334/82-01 RESULTS:
PAGE 2 of 10 SCOPE: (Cont.)
stud and three washers in the number two cylinder of the diesel generator that had been furnished to the Grand Gulf Nuclear Generating Station, Unit No. 1.
The Engine and Compressor Division of Transamerica Delaval, Inc., had also issued 10 CFR Part 21 reports on the aforementioned items, as well as on a potential problem with a diesel generator starting air check valve.
PLANT SITE APPLICABILITY: (Cont.) 50-329, 50-330; 50-445, 50-446; 50-440, 50-441;* 50-438, 50-439;* 50-518, 50-519, 50-520, 50-521;* 50-553, 50-554.*
- Applicable plant unit not defined.
A.
VIOLATIONS:
None B.
NONCONf0RMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 16.7.1 of the QA Manual, QA Records were not being protected against fire, damage, or loss.
I 2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 7.3.1 of Section 7 of the QA Manual, a welder was observed welding without the applicable welding procedure being either in his possession, or having been issued in the job package as required by instructions.
3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Specifi-cation 100-A-3 and corrective action commitments, weld material was not returned to stores within the prescribed time period.
4.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Transamerica Delaval, Inc., commitment as stated in their letter of June 5,1981, to NRC, the audit of the ASME Weld Shop had not been performed as committed.
5.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 4.4.3 of Section 4, dated February 27, 1981, of the Quality Assurance Manual, Betts Spring Company, a supplier of valve springs (major /
critical), had not been surveyed a minimum of once every three years to assure their ability to comply with the specification requirements and to review the implementation of their quality program.
This was evidenced by a Vendor Quality Program Survey (VQPS) form which had been completed February 25, 1976.
An updated VQPS was received by TDI-ECD on January 27, 1982.
This form is completed by the vendor and does not assure their ability to comply with specifications or review implementation of their quality program.
152
ORGANIZATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND, CALIFORNIA REPORI INSPECTION NO.
99900334/82-01 RESULTS:
PAGE 3 of 10 6.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 6.1.3 of the Approved Supplier List (ASL) procedure, dated March 20, 1980, Associated Spring Company (Barnes Group) had been placed on the ASL, dated August 1981, prior to completion of a survey or audit as evidenced by an incomplete Audit Form, dated August 12, 1981.
Additionally, a purchase order, dated October 21, 1981, ordered valve springs.
C.
UNRESOLVED ITEMS:
1.
The ASL, dated August, 1981, lists Associated Spring Company (Barnes Group), a new supplier, with the highest possible rating (10.0);
however, no hardware had been received from them.
The ASL procedure indicates that a supplier is rated according to the disposition of received hardware but it does not address how a numerical rating will be awarded to a new supplier.
As a result of the foregoing, the inspector was unable to determine the validity of awarding a perfect rating to an unknown quantity; i.e.,
a vendor who has supplied no hardware, therefore, no history of received hardware.
2.
The manufacturer (Melrose Spring and Tool Works) of defective valve springs had been excluded from the ASL, dated August 1981, as a source of supply for the valve springs.
That action precludes receipt of defective springs from that source; however, the manu-facturer was not performing a unique operation.
There was no indi-cation that Delaval had identified the cause of the problem.
Based on the foregoing, it was nc'. apparent that adequate measures had been taken to preclude recurrence.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Deviation A (79-01) - Procedures and procedure manuals had not been established for the manufacturing and material control departments.
The inspector verified by reviewing revised procedure manuals that the manuals now identify all departments that will main-tain procedures and procedure manuals.
The corrective action was taken prior to the stated completion date.
2.
(Closed) Deviation B (79-01) - The vendor was not recording actual readings of measuring and test equipment undergoing periodic recali-bration.
The inspector verified by review of written instructions issued to supervision, and review of the revised IP 100 of the QA 153
ORGANIZATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND, CALIFORNIA REPUkT IN3Pttilua NO.
99900334/82-01 RESULTS:
PAGE 4 of 10 Manual, that corrective action and preventive measures have been taken to prevent recurrence.
The action was taken prior to the committed completion date.
3.
(Closed) Deviation C (79-01) - The Qualified Suppliers List had not been updated monthly nor had a monthly summary of the quality rating of vendors been forwarded to QC and Purchasing.
The inspector reviewed the revised issue of IP 700 relating to the updating and issuance of the approved vendor summary to purchasing and other designated departments.
Current lists and summaries indi-cate full compliance with the revised IP 700 section of the QA Manual.
4.
(Closed) Deviation D (79-01) - The activities, use and/or status of Corrective Action Requests (CAR) had not been reported by the Manager of Quality Control to the Division General Manager on a quarterly basis.
The inspector verified that the Quality Control Manual had been revised to control the issuance of CAR activities to the Division General Manager.
The new reporting requirements are contained in the QA and ASME Manuals.
Section 14-1 and 15-1.
The procedures are currently being implemented as required.
5.
(0 pen) Deviation E (79-01) - The Document Control Center has not been protected against fire.
The inspector verified that the vendor had purchased new cabinets and had transferred records to a microfilm system.
During the inspector's review of the corrective action taken as a result of this deviation, it was noted that the current QA Manual Section 16.7.1, dated February 27, 1981, still requires the records to be protected from fire at Delaval.
In addition, the Storage Cabinets for the microfilm are not rated as fireproof.
(See Nonconformance B-1).
6.
(Closed) Deviation F (79-01) - Audits of the Division Quality Assurance programs processes and procedures were not performed semiannually by Quality Engineering.
The inspector verified that the Quality Assurance Manual, Section 17, dated February 27, 1981, had been revised to prevent recurrence of this item.
Audits are now per-formed annually.
Completed audits for 1980-1981 were reviewed for compliance.
7.
(Closed) Deviation G (79-01) - Industrial Engineering had not reviewed, approved or disapproved Engineering Change Notices, Engineering Log not maintained and Drawing Change Requests had not been classified as major or minor.
The inspector reviewed the corrective action taken by TDI to prevent recurrence.
The Engineering Division Standard Prac-154
ORGANIZATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND, CALIFORNIA RLPORI IN5PELi10N NO.
99900334/82-01 RESULTS:
PAGE 5 of 10 tice Bulletin has been revised and reissued.
The Engineering Change Notice Log had been revised and was being filled out as required.
Current Change Notices were reviewed for compliance and were found to be acceptable.
Compliance to the required changes were completed prior to the stated completion date.
8.
(Closed) Unresolved Item (79-01) - %uthority and dates of manufacturing and material control personnel performing activities affecting safety related functions had not been clearly established and delineated.
The inspector verified that the current QA Manual clearly delineates the authority and duties of all personnel performing safety related activities.
9.
(Closed) Deviation B (80-01) - Quality Engineering did not process a Corrective Action Request Form (CAR) for identified TDI failures to meet weld quality contract requirements in Diesel Engine Generator piping.
The inspector verified by review of past and current issued corrective action requests that quality engineering is now generating CAR's as required by QA and contract requirements.
In addition, the inspector reviewed documentation pertaining to the meeting held with all welding personnel on April 6,1979.
The corrective action was taken prior to the stated completion date.
10.
(Closed) Deviation C (80-01) - TDI written practice IP 600 did not describe procedure for examination of Level III personnel.
- Also, personnel performing visual examinations had not been qualified to methods covered by SNT-TC-1A documents.
The inspector verified by review of documentation that corrective action had been taken to prevent recurrence of this item.
IP 600 has been revised and persons performing visual inspection have been requalified.
Training sessions were held and documented for all personnel.
Internal audits had been conducted to prevent recurrence.
The corrective action had been taken prior to the stated completion date.
11.
(Closed) Deviation D (80-01) - Welding was being performed by a welder in the 3G position who had been qualified for the 1G position only.
The inspector verified by review of welder qualification records that all welders had been recertified to the 6G position.
The current qualified welders were reviewed for compliance.
Current welder certifications have been reviewed by the committee appointed for the task.
Corrective action had been taken prior to the stated completion date.
12.
(0 pen) Deviation E (80-01) - Unused weld rod was not being returned to the storage area within four hours of issuance.
The inspector verified by review of documentation that all welders were notified of the required control of weld materials.
Corrective Action Requests 155
ORGANIZAIION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND, CAL r0RNIA ktFuki taaettiiva i
N0.
99900334/82-01 RESULTS:
PAGE 6 of 10 were issued and control of weld materials were verified during an audit performed on March 27. 1981. A further instance was identified during this inspection of failure to return rods within four hours of issuance.
13.
(Closed) Deviation F (80-01) - Welder identification and verification of welders not identified on Route Sheets.
The inspector verified by review of Production Routing Sheets that welders are being properly identified and documented on Route Sheets.
A CAR was issued and training sessions have been held.
Internal audits have been performed of welding operations.
14.
(Closed) Deviation G (80-01) - Defective weld removed and replaced without rejection and documentation on Inspection Report.
The inspector reviewed documentation and corrective action taken to resolve this item.
A CAR was issued to cover the repair - an ASME rework Record has been developed and is in use.
Audits have been conducted.
i 15.
(Closed) Deviation H (80-01) - Weld was observed to contain an area with less than a required h inch fillet resulting from a fit-up condition.
The inspector reviewed current inspection reports for compliance.
A CAR was issued covering the above deviation.
Training for NDE personnel was conducted and eye examinations were given to all NDE personnel.
16.
(Closed) Deviation I (80-01) - Inspection acceptance stamps had not been entered for operations on the production sheet, nor had inspection reports been prepared to denote a rejected condition after inspection.
The inspector verified that a CAR was issued to cover this item.
Current jobs in progress were reviewed for compliance to IP 500 require-ments.
Audits have been conducted for compliance.
17.
(Closed) Unresolved Item B.4.d (80-01) - Discrepancies in dates that seismic tests to battery rack were conducted prevented a definite acceptance of the Seismic Report and Certification.
The inspector verified that a corrective revision to the Seismic Analysis Report had been issued.
The Seismic Test was run from January 1, 1979 to July 19, 1979.
The original package was dated March 1979.
A corrected certificate now indicates the correct date.
18.
(Closed) Unresolved Item D.4.c (80-01) - Rework Tag No. R-0959 did not indicate by inspection if the work had been performed.
The inspector reviewed documentation that indicated that the requieaments of IP 300 QA Section 14 are currently being followed.
Rework tags are now being properly filled out.
156
ORGANIZATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION OAKLAND. CALIFORNIA REPORI INSPECllON NO.
99900334/82-01 RESULTS:
PAGE 7 of 10 19.
(Closed) Nonconformance A (80-01) - Welders using weld rod not approved for weld procedure.
The inspector reviewed the corrective action taken by TJI to prevent recurrence.
Training sessions were held with all welders.
Each welder was issued his personal copy of weld pro-cedures.
The inspector verified that current weld procedures have been rewritten on a short form.
These forms would be issued with the job package and used on all future jobs.
The procedure would be in the possession of the welder.
(See nonconformance 8.2).
20.
(0 pen) Nonconformance B (81-01) - Specification 100-A-3 requires all weld rod be returned to stores after four hours.
The inspector reviewed the corrective action taken by TDI to prevent recurrence.
Training sessions were held with welding personnel regarding return of weld material before the time limit had been exceeded.
In addition, the weld issue room attendant was instructed to monitor the time out of weld rod.
He would contact welders on a three hour schedule to prevent running over the four hour schedule.
(See nonconformance 8.3).
21.
(Closed) Nonconformance C (81-01) - Welder observed welding out of procedure electrical parameters.
The inspector verified that a CAR had been issued covering the item.
A training session for all welders was held and documented.
Welders were instructed in the use of welding procedures.
22.
(Closed) Nonconformance D (81-01) - An identified nonconforming compo-nent had not been documented on the inspection report form or placed in bond.
The inspector verified that corrective action reports had been prepared by TDI inspection personnel.
A meeting with QC staff was held and documented.
Current nonConformances have been documented and when required, placed in segregated areas.
The audit of the weld department scheduled for the fall of 1981 had not been performed.
(See noncon-formance B.4).
23.
(Closed) Nonconformance A (81-02) - The parts list and component drawings released by Engineering had not defined the acceptance criteria for the oil plugs.
In addition, the installation instruction for the oil plugs contained no acceptance criteria.
The inspector verified thru the review of revised drawings, corrective action requests, docu-mentation of training of industrial engineering and engineering person-nel, and a revised Route Sheet for the oil plug that corrective action, as stated in the TDI letter of October 6, 1981, was acceptable.
The steps taken to prevent recurrence were considered acceptable.
157
ORGANI7ATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION HLPORi IN3ettilua NO.
99900334/82-01 RESULTS:
PAGE 8 of 10 24.
(Closed) Nonconformance B (81-02) - Measures were not established to assure that tools used in the crankshaft oil plug installation are properly controlled and adjusted at specified periods to maintain accuracy within the necessary limits.
The inspector reviewed the procedure and equipment used to install the oil plug.
The tool used is a rolling type tool which is opened by use of hydraulic pressure.
The final proof of the rolling operation is the leak test that is applied to the installed plug.
The inspector reevaluated the nonconformance and determined that the tool is being controlled in a manner consistent with the requirements of 10 CFR Part 50 and the TDI QA Manual requirements.
25.
(0 pen) Nonconformance C (81-02) - Records had not been maintained to furnish evidence that the motors for the auxiliary lube oil and jacket water pumps had been environmentally qualified.
This item remains open for further NRC evaluation.
26.
(0 pen) Nonconformance D (81-02) - Documentary evidence was not available to assure that the seller of the motors, for the auxiliary lube oil and jacket water pump had complied with the requirements of the purchase order.
This item remains open for further NRC evaluation.
l 27.
(Closed) Unresolved Item C.1 - Assembly / test Route Sheet inspection elements not signed off.
The inspector reviewed the Route Sheets currently in use and found that they are being signed off as required by instructions.
The current Route Sheets have been r^ vised to prevent the recurrence of the above item.
Acceptable corrective action and l
preventive measures have been taken.
E.
OTHER FINDINGS OR COMMENTS 1.
Follow up on 10 CFR Part 21 Reports a.
Potentially Defective Valve Springs (Associated, Butts and Connor) - Currently, there are three sources for valve springs, Part No. 03-360-02-0M.
Purchase orders had been placed with Associated and Betts.
A review of three vendor inspection reports (Form P-313), dated from February 24, 1978, through January 5, 1982, revealed that valve springs of the specific part number had not been received from Betts in 1979 and 1980.
An unidentified, undated document indicated that TDI-ECD visited l
1 l
.,. ~
158
ORGANIZATION:
TRANSAMERICA DELAVAL, INC.
ENGINE AND COMPRESSOR DIVISION O AK L Af,1, CALIFORNIA REPORI INSPECIION NO.-
99900334/82-01 RESULTS:
PAGE 9 of 10 Betts on July 24, 1981, for a plant tour.
Prior to the close of this inspection, a completed Vendor Quality Program Survey form from Betts was received by TDI-ECD.
This form was completed by Betts and does not assure their ability to comply with specifi-cations or review implementation of their quality program.
Associated was identified as a new supplier who had supplied no hardware.
While the supplier of the defective springs (Melrose) had been excluded from the ASLNCSRE, dated August 1981, there was no apparent indication that TDI-ECD had taken adequate measures to assure that valve springs from current suppliers would not be defective.
b.
Inadequate Starting Air Check Valve - The starting air check valve leaked during a seismic test at Structural Dynamic Research Corporation.
The goal was a leakage rate of zero, however, the manufacturer (Wm. Powell Valve Co.) deemed that an unrealistic goal.
Four valves had been returned to The Wm. Powell Valve Co., for rework and testing to the TCI-ECD newly established leak rate.
Investigation is still underway; TDI-ECD submitted a revised 10 CFR Part 21 report on January 27, 1982.
c.
Cracked Stud and Spherical Washers - The use of spherical washers in the piston assembly was discontinued in late 1976 to mid-1977.
At that time, a new design using belleville washers was implemented.
On December 31, 1980, TDI-ECD Service Information Memo No. 324 (among others) was distributed to all users of Model R and RV diesel engines below Serial Number 75080.
That document provided instructions for modifying the piston assembly and replacing the spherical with belleville washers.
Several standby diesel generators have successfully undergone qualification tests subsequent to implementation of the design change.
d.
Paragraph 21.21(b)(3), and its subparagraphs (vii) and (viii) of litle 10, Part 21, of the Code of Federal Regulations requires that the written report contain to the extent known:
(1) correc-tive action which had been, is being, or will be taken; and (2) any advice related to the defect that had been, is being, or will be given to purchasers or licensees.
159
ORGANIIATION:
TRANSAMERICA DELAVAl, INC.
ENGINE AND COMPRESSOR DIVISION DAKLAND, CALIFORNIA RtPORI INM t1iluN NO.
99900334/82-01 RESULIS:
PAGE 10 of 1C The Transamerica Delaval, Inc. - Engine and Compressor Division's (IDI-ECD) 10 CFR Part 21 report, dated July 30, 1981, related to potentially defective valve springs, did not include that total (1) corrective action, and (2) advice given to purchasers or licensees.
Specifically, (1) the supplier of the potentially defective valve springs was removed trom the Approved Suppliers List, and (2) a letter to purchasers or licensees provided advice on recognition of the suspect valve springs.
2.
Follow Up On A Regional Request Mississippi Power and Light Company's 10 CFR Part 50.SS(e) report, dated December 10, 1981, addressed the item discussed above (paragraph D.1. c ).
Additionally, the report identified the following deficiencies:
(1) missing wrist pin keeper, (2) grooves in cylinder liners; and (3) excessive bearing wear and deterioration.
TDI-ECD informed the inspector that these had been discussed during a Delivered Product Trouble meeting and determined to be workmanship problems.
No docu-ments were presented to substantiate that conclusion.
160
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT IN5PECIION 4/5-8/82 INSPECTION NO.
99900510/82-02 DATE(S) 4/19-23/82 ON-SITE HOURS: 86 CORRESPONDENCE ADDRESS:
United Engineers and Constructors, Inc.
j ATTN:
Mr. R. A. Curnane, Vice President Project Support Operations 30 South 17th Street Philadelphia, PA 19101 l
l ORGANIZATIONAL CONTACT:
Mr. J. B. Silverwood, QA Manager l
TELEPHONE NUMBER:
(215) 422-4744 PRINCIPAL PRODUCT: Architect Engineering Services NUCLEAR INDUSTRY ACTIVITY: United Engineers and Constructors (UE&C) is the Architect Engineer on the following major projects:
Seabrook, Units 1 and 2, and Washington Public Power Supply System, Unit 1 (WNP-1).
Also, UE&C has active engineering service contracts on two nuclear plants for two utility clients.
SfT//h ASSIGNED INSPECTOR:
D. D.-Chamberlain,/ Reactor Systems Section (RSS)
Dhte 8-OTHER INSPECTOR (5): J. Conway, Reactive & Compenent Program Section S. K. Chaudhary, Region I APPROVED BY:
(
/[
C. de, pal 6, Chiel,'RSS Jate '
INSPECTION BASES AND SCOPE:
A.
BASES:
UE&C Topical Report UEC-TR-001-5A, PSAR for the WNP-1 project and
(
B.
SCOPE:
Status of previous inspection findings, follow up on items from the WNP-1 site inspection, inspection of UE&C Seabrook site design activities
(
(interface between site design group and home office design group),
(
and inspection regarding the following requests:
(1) Region I request relating to a 50.55(e) report on WNP-1 regarding pipe support design errors and subsequent determination of similar errors on the (cont. on next page)
PLANT SITE APPLICABILITY:
This inspection relates to the following plant dockets: 50-443/444, and 50-460.
1 61
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.
99900510/82-02 RESULTS:
PAGE 2 of 10 SCOPE: (Cont.) Seabrook project; (2) Region I request relating to UL&C audit of Grinnell Fire Protection System Company; and (3) 10 CFR Part 21 report for failure of UE&C to consider LOCA temperature effects in design of platform structural steel framing within containment s
(WNP-1).
A.
VIOLATIONS:
None 6.
NONCONFORMANCES:
/
1.
Contrary to Appendix II of Topical Report UEC-TR-001-5A and Section 8.2 of ANSI N45.2.11, Administrative Procedure No. 15, Revision 14, Changes to Project Documents, allows " minor" design changes to be made without requiring the same review and approval as the original design document (Seabrook).
2.
Contrary to Section 17.1.3.1 of Topical Report UEC-TR-001-5A and sections III and V of procedure GEDP-5, provisions were not being implemented for the preparation, review, and approval of a calculation on conduit and cable tray support weld adequacy (Conduit Support Calculation dated April 25, 1980).
3.
Contrary to Criterion V of 10 CFR Part 50, Appendix B and Project Procedure No. 34, Project Change Requests after January 1981, and all Project Change Proposals for the WPPSS Nuclear Project No. 1 were not microfilmed.
C.
UNRESOLVED ITE_MS:
It was not apparent that the present tracking system for incorporation of design changes into design documents will assure that the required time limit (60 days) of procedure AP-15 will be adhered to.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (81-03):
The storage of QA records for the Seabrook Project was not meeting the intent of the storage requirements of ANSI N45.2.9, and audits of record storage areas were not being performed on a periodic basis.
162
ORGANIZATION:
UNITED ENGINEERS AND CON 5TRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.
99900510/82-02 RESULTS:
PAGE 3 of 10 The inspector selected three design drawings and three specifications from the Document Control Center System's Report and the Engineering-Purchasing Schedule, respectively.
The inspector verified that each document was microfilmed and one copy 16 MM jackets for the specs and 35 MM aperture cards for the drawings, along with a hard copy are retained in file cabinets in the Document Control Center (DCC) located on the sixth floor of the UE&C building.
A second copy of the microfilm is retained in file cabinets in the Record Retention Room (RRR) on the fourth floor.
The QAB of NRR has recently approved (ref. Haass/Ebiner letter, dated April 26, 1982) UE&C's dual storage facilities as meeting the require-ments of ANSI N45.2.9, and this inspection has verified the satisfactory implementation of the storage requirements.
Audit report NH-513,
" Audit of Storage of QA Records," conducted on March 23-24, 1982, at the home office, the Boston office, Seabrook site, and Yankee Atomic Electric Company (Framingham, MA), was reviewed.
This report satisfies the inspector's earlier concern and meets UE&C's commitment to audit quality-affecting activities on a yearly basis.
In inspecting the QA record activities relating to Seabrook, two instances of missing documents were noted: a hard copy of a specifi-cation was not in the file (DCC) and the date of removal and the individual who had removed it were not identified on the Document Custody Record Sheet; an aperture card of the latest revision of a drawing was not in the master aperture card file (RRR), and there was not documentation that the card had been removed.
This area will be considered further during subsequent inspections.
2.
(Closed) Nonconformances (81-03):
The vendor surveillance file for two contracts on the Seabrook Project revealed that some witness points were not being accomplished or waived with justification.
Based on these findings, UE&C committed to review all safety-related orders for Seabrook to complete the required vendor surveillance The inspector verified that'the Vendor Surveillance Branch of the Reliability & QA Department has reviewed the Seabrook contracts and identified those containing incomplete surveillance points.
Certifi-cation of witness points on some of the questionable contracts have been completed.
The remaining contracts will be completed in the future.
Verification that these activities have been completed will be accomplished on a future inspection.
163
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.
99900510/82-02 RESULTS:
PAGE 4 of 10 3.
(0 pen) Nonconformance (82-01):
The Reliability and Quality Assurance Department did not assure that the committed completion date was met for corrective action in response to a previous NRC inspection f inding (i tem B.1, report 81-03).
l This item will remain open pending the review of UE&C response to the 82-01 inspection and review of corrective action.
The response was not received in time to review prior to the 82-02 inspection.
4.
(0 pen) Nonconf ormance (82-01):
The transferring organization for three pipe stress analysis packages examined (WNP-1) did not verif y that the latest revision of the Composite / System Piping Drawing and Fabrication and Erection Drawing was referenced on the analysis drawing.
This item will remain open pending the review of UE&C response to the 82-01 inspection and review of corrective action.
The response was not received in time to review prior to the 82-02 inspection.
5.
(Closed) Unresolved Item (82-01):
It was not apparent that program requirements were adequately defined and/or being followed in the preparation of nuclear pipe support design packages for transfer of responsibility to the WNP-1 site design group.
Pipe Support Guideline PSG-1-1004, Nuclear Pipe Support Closeout Program, has been implemented and the requirements were being followed on nine pipe support packages examined.
E.
OTHER FINDINGS OR COMMENTS:
1.
WNP-1 Site Inspection Followup - As a result of the WNP-1 site inspection conducted on February 2-5, 1982, certain areas were identified for followup action at UE&C home office. lhe results of the followup action for each item is as stated below:
a.
The implementation of two new procedures (PP-47 and PSG-1-1004) were to be monitored during future NRC inspections.
PSG-1-1004, Nuclear Pipe Support Closcout Program, has been implemented (D.5 above).
PP-47, Closecut and Transfer of Design Documents, has not been implemented and UE&C management stated that they do not feel that PP-47 will be required.
The NRC inspector examined two internal memoranda and field procedure FGCP-28, which address transfer of design documents to the field design group.
An additional internal memorandum (AAC0259) was written during this inspection which describes the requirement for a receipt acknowledgement f or any documents which transfer respensibility for continuing engineering from the home office 1 64 l
l
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORI INSPECTION L
N0.
99900510/82-01 RESULTS:
PAGE 5 of 10 to field engineering.
Each supervisory discipline engineer ir required to maintain up-to-date records as to what responsi-bilities/ documents have been transferred to the field.
It was noted that FGCP-28 does not exactly match the present field organizational structure and UE&C management stated that a revision to FGCP-28 is in process.
The review of the revised FGCP-28 will be referred to Region V for followup.
b.
The backup calculations and documentation of verification of the cook book design document for cable tray support design and HVAC duct support design were examined, and the program requirements were being implemented.
The design review records for six System Design Descriptions c.
(SDD's) were examined and checked against certain design documents.
Progrv s eguirements for verification of SDD's were being imple.nented.
2.
UE&C Seabrook Site Design Activities - This area of inspection was conducted as a result of a request from Region I to review UE&C Seabrook site design activities and the interface between UE&C site design and UE&C home office design.
An inspection was conducted at the Seabrook site on April 5-8, 1982, followed by an inspection at UE&C Philadelphia office on April 19-23, 1982.
The areas examined and results are discussed below:
Seabrook Site - The UE&C site engineering organizational structure a.
and applicable procedures for control of design changes /documen-tation were examined.
The organization and the design change procedure (AP-15) were revised early this year to provide better control of design changes and to provide faster response to con-struction problems.
Twenty-two design change documents were examined to verify implementation of program requirements.
A review was made of the UE&C interface with contractors in the HVAC duct support and the cable tray support areas.
Two poten-tial problem areas were referred to the Seabrook resident inspector for followup:
(1) administrative problems with structural calculations performed on site; and (2) question of the adequacy of as-built information to be provided by the electrical contractor for cable tray supports.
l 165
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.
99900510/82-01 RESULTS:
PAGE 6 of 10 Certain other items were identified for follow up at the UE&C Philadelphia office.
fhe items /results are discussed in the following section.
b.
UE&C Philadelphia Office - The follow up of the Seabrook site inspection items resulted in two nonconformances (B.1 and 2 above) and one unresolved item (C. above).
The results of of the followup action are as stated below:
(1)
The instructions for control / distribution of the change control log for tracking of design changes were reviewed and the control / distribution meets program requirements.
(2) The site /home office interface for control / distribution of Engineering Change Authorizations (ECA's) was reviewed.
The distribution requirements are listed on the back of the ECA form.
Every discipline gets a copy of all ECA's, and the responsible discipline for home office concurrence / action is identified by a code number which is a part of the ECA number.
fhis area meets program requirements.
(3)
The old program for control of field design changes used Interim Change Authorizations (ICA's) which allowed work to begin prior to completion of all analysis.
Some of the ICA's indicated " MAG ANALYSIS" required, and this requirement was tracked by the change control log.
The review of this area revealed that there are approximately 25 open ICA's that require " MAG ANALYSIS."
The status of these ICA's will be reviewed during future inspections.
(4) The revised procedure for design change control (AP-15) requires " major" changes that are approved by the site design group to receive home office concurrence or no con-currence within 10 days.
An attempt was made to determine how this will be controlled / tracked / accomplished by the home office design group and how the site design group is aware of concurrence status.
UE&C management stated that this area is still being reviewed and that AP-15 would be revised to more clearly define responsibilities.
This area will be reviewed turther during a future inspection.
166
l l
l l
ORGAN!/ATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA RLP0HI IN5PLLl10N NO.
99900510/82-01 RESL'LTS:
PAGE 7 of 10 (S) During the review of HVAC duct support ECA's at the site, it was determined that a structural review of ECA's involving attachment; to structural steel had not been required.
The site group stated that the " Beam Verification Program" at the home office would pick up all attachments to structural steel.
A review was conducted of the planned " Beam Verifi-cation irogram," and the stated purpose of the program was to analyze each structural steel beam first with as-designed loading and later with as-built loads.
The as-built program must be detailed enough to provide locations of attachments to structural steel.
This program is just in the planning stages, and the program implementation will be fellowed during future inspections.
(6) The control of the 60-day, 3-ECA limit requirement of AP-15 for incorporation of design changes into design documents was reviewed and UE&C has a computer log (R) which tracks the number of outstanding changes by discipline.
It was not clear how this tracking system will assure that changes are incorporated within the required time frame.
This item is unresolved (C. above).
(7) Procedure AP-15, Changes to Project Documents, classifies changes as " major" or " minor." Major design changes require home office concurrence, but minor design changes only require site approval.
A list was prepared by the site design group to identify the types of design changes that are " minor changes."
The list of minor charges had no review and approval requirements and minor change ECA's could be prepared and approved by the responsible site engineer only.
The area of " minor" design changes was identified as a nonconformance (B.1 above).
(8) A review was made of the verification of the cook book design document for cable tray support design.
A nonconformance was identified relating to a calculation f or weld sizing of cable tray support connections to structures (B.2 above).
(9)
The site design group may produce preliminary calculations for their approval of ECA's, and these calculations must be f orwarded to the home of fice for concurrence.
If the home office uses the calculations performed by the site design group, they must assure that all program requirements 167
ORGANI/ATION:
UNITEC ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.
99900510/82-02 RESULTS:
PAGE 8 of 10 for calculations are satisfied.
A Region I inspector accompanied the VPB inspector during this inspection, and the Region I inspector reviewed the area of site calculations for ECA's.
It was determined that presently no site calculations have been used for home office concurrence of ECA's.
3.
Pipe Support Design Errors (WNP-1/Seabrook) - This area of inspection resulted from a Region I request for follow up on a 10 CFR Part 50.55(e) report from WNP-1 regarding pipe support design errors and the sub-l sequent determination of similar errors on the Seabrook project.
The types of errors identified were minor human errors, failure to follow code requirements (e.g., weld sizing), failure to pick up change in design criteria of supplier, etc.
The action taken by UE&C beginning in early 1980 includes:
a.
Generic:
(1) UE&C issued formal size support design guidelines for use by all design personnel.
(2) Formal checklists were instituted to be used for verifica-tion that critical design areas had been considered for individual pipe support designs.
(3) Additional training was conducted for all pipe support design personnel.
I b.
WNP-1 Project:
(1) UE&C reviewed approximately 5,000 pipe supports which l
revealed about 1,700 with errors.
(2) The 1,700 with errors were put on a " soft hold" which means that they can be installed but cannot be signed off.
(3) The "sof t hold" hangers are being given priority for release under the nuclear pipe support transfer program which corrects the errors identified.
168
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA l
l REPORT INSPECTION NO.
99900510/82-01 RESULTS:
PAGE 9 of 10 c.
Seabrook Project:
(1) UE&C sampled approximately 85 pipe supports out of 1,500 designed prior to February 1, 1980.
The sampling revealed 24 with errors.
(2) A purchase order is in the process of being issued for a totally independent review of the remaining 1,500 pipe supports.
Although the action taken by UE&C, to date, appears adequate, due to time constraints during this inspection, the NRC inspector was not able to verify certain areas such as the training conducted, sampling of design performed after early 1980, other project work affected, etc.
Therefore, this item will remain open pending review during a future inspection.
4.
UE&C Audit findings of Grinnell (Seabrook) - An examination of UE&C's external audit activities indicates that the Audit Branch and Project Quality (Boston office) of the Reliability & QA Department conducted a facility survey (March 1981) and three audits (July 1981 through March 1982) of Grinnell at their Providence, Rhode Island, office and their Cleveland-Rowen, North Carolina, fabrication plant.
An evalution of these audit findings indicates satisfactory imple-mentation of Grinnell's QA program for fabricating items for fire protection systems at the Seabrook facility.
5.
10 CFR Part 21 Report (WNP-1 and 4) - UE&C did not use the current design LOCA temperatures as they were not considered a design input requirement at the time the calculation for the structural steel supports were performed.
Following the issuance of a 10 CFR Part 21 report in December 1981 on the above subject, Region V became concerned about a potential design change control problem at UE&C.
urui,.g this inspection, several procedures addressing design control were reviewed to determine UE&C's interface requirements between the home office engineering group and the site engineering group.
For the WPPSS project, the site has the tracing for each drawing and can make design changes without home office approval.
However, for the steel supports within containment, the home office was notified in writing of the connection changes made in the field for the embedded plates.
These changes were utilized in the home office's updated calculations (i.e., considered LOCA temperature effects) for the steel supports.
A review of several calculations indicated that an independent review was performed on each calculation.
Proposed changes on some of the supports are also being verified by an independent laboratory (Lehigh Testing Company).
,y 169
ORGANIZATION:
UNITED ENGINEERS AND CONSTRUCTORS, INC.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.
99900510/82-01 RESULTS:
PAGE 10 of 10 Several Requests for Information (RFI), which include the calculations and subsequent changes to drawings, were sent to the site to update the applicable tracing.
The revised drawing is microfilmed, and copies of the drawing and the microfilm are sent to the home office.
Project Change Proposals (PCP) incorporating the appropriate RFI's were written.
A PCP requires concurrence by the licensee prior to making any changes in the field.
For the Seabrook project, the " minor" changes that the field may make without home office approval are documented.
Unlike the WPPSS project, the tracing of each drawing is retained in the home office.
l 170
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION ELECTRICAL SYSTEMS DIVISION HUNT VALLEY, MARYLAND ktP0ki INSPECTION INSPECTION NO.
99900217/82-01 DATE(S) 4/19-22/82 ON-SITE HOURS: 24 CORRESPONDENCE ADDRESS:
Westinghouse Electric Corporation Electrical Systems Division ATIN:
Mr. J. F. Heins, General Manager 1111 Schilling Road Hunt Valley, Maryland 21031 ORGANIZATIONAL CONTACT:
Mr. R. W.
Lee, Product Assurance Manager TELEPHONE NUMBER:
(301) 667-5532 PRINCIPAL PRODUCT: Electronic components and controls NUCLEAR INDUSTRY ACTIVITY: About 75% of the manufacturing is commercial nuclear work.
The product lines include Solid State Protection Systems, Nuclear Instrumentation Systems, Digital Rod Position Indication Systems, Reactor and Containment Vessel Level Monitoring Systems, and Technical Support Centers.
b f/!82-ASSIGNED INSPECTOR:
W.
M McNeill, Reactive & Components Programs Da'te Section (R& CPS)
OTHER INSPECTOR (S):
APPROVED BY:
.~4 s//d?
I. Barnes, Chief, R& CPS Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
Manufacturing process control, control of special processes, audits, and status of previous inspection findings.
PLANT SIrr APPLICABILITY:
Not Identified.
171
ORGANilATION:
WESTINGHOUSE ELECTRIC CORPORATION ELECTRICAL SYSTEMS DIVISION HUNT VALLEY, MARYLAND RFPORT INSPECi10N i
NO.
99900217/82-01 RESULTS:
PAGE 2 of 4 A.
VIOLATIONS:
None B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, and section 9.0 of the Product Assurance Manual, no Quality Assurance Procedure was established which addresses the qualification of the wire wrap process ard personnel.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and section 2.1 of Quality Assurance Procedure (QAP) 10.2, one of two operators performina Termi point wiring on April 20, 1982, of a cabinet assembly 106E58G02, SN 0002, was found not to be certified in accordance with the procedure.
No records could be found to demonstrate that the required training and testing had been accomplished.
3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and section 2.2.1 of QAP 13.3, although wire wrap was observed in the PC Board area on several boards, the applicable Process Specification (PS 82355 HA) was not maintained at the Inspection Station. The speci-fication identified the acceptance criteria to be used by inspection.
4.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and section 3.1.1.1.(9) of QAP 14.6, Requests for Engineering Action (REA's) 342565 and 342566, were observed on serialized equipment which did not have the assigned serial numbers entered on the REA's.
5.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and section 2 of QAP 14.1, the procedure was found to be not fully implemented in that:
a.
The requirements of paragraph 3.1.1.3.13 of the procedure, per-taining to updating of tags, was not accomplished for Job No.
10477, part 1064E03, as of April 21, 1982, to reflect change order RN3088, which had been issued on March 23, 1982.
b.
Recording of all operators signatures and dates, which is a requirement of paragraph 3.2.1.1. of the procedure, was not accomplished on the tags for parts 1063E43G41, SN0001, and 1063E43G39, SN0001, for Job No. 10477.
172
ORGANIZATION:
WESTINGHOUSE ELECTRIC CORPORATION ELECTRICAL SYSTEMS DIVISION HUNT VALLEY, MARYLAND REPORT INSPECTION NO.
99900217/82-01 RESULTS:
PAGE 3 of 4 Shortages (i.e., missing serial number stamping and missing c.
brackets, and three power supplies) were not identified on the tag, as required by paragraph 3.2.1.4 of the procedure, for part 1063E43G39, SN0001 for Job No. 10477.
d.
A Defective Article Report (DAR) number was not entered on the tags for part 6090072G01, SN0005 foi Job No. 10531, although required by paragraph 3.3.1.6.1 of the procedure.
It was also noted that two nonapplicable REA's had been recorded on the tags.
C.
UNRESOLVED ITEMS None D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (81-02):
Two Inspection Control Tags were found with incorrect and missing information.
The inspector verified that the tags in question had been corrected and the departmental meetings were held as a preventive measure.
This area was reinspected during this inspection and a nonconformance identified (see B.5).
2.
(Closed) Nonconformance (81-02):
Two Corrective Action Requests (CARS) for vendor nonconformances were either not initiated or were voided.
A review of current MRB activities found no further problems in this The corrective action on the previous problems was implemented area.
and an audit of the vendor is underway.
3.
(Closed) Nonconformance (81-02):
Three cases were found where defects were not logged or were logged incorrectly.
An inspection of current Inspection Submittal and Discrepancy Records and nonconforming hardware found all defects properly logged.
Training sessions had been held on this problem.
4.
(Closed) Nonconformance (81-02):
Independent verification of vendor test results has not been performed for the last 4 months.
A review of the receiving inspection test report log found the procedure in question to be implemented.
Internal memos were issued to implement the pro-cedure requirements for independent verification.
173
ORGANIZATION:
WESTINGHOUSE mLECTRIC CORPORATION ELECTRICAL SYSTEMS DIVISION HUNT VALLEY, MARYLAND RLPORI IN5PEC110N NO.
99900217/82-01 RESULTS:
PAGE 4 of 4 E.
OTHER FINDINGS OR COMMENTS:
1.
Manufacturing Process Control - The Inspection Control Tag procedure was reviewed and its implementation inspected.
A sample of six parts and thei associated tags were inspected in the PC board and wire assembly areas.
In this area nonconformances B.4 and B.5 were identified.
In regard to nonconformance B.4 it was noted that engineering had a date entered on the REA where a serial number should have been entered.
In regard to nonconformance B.5 it was noted that the inspection of these units had had yet to be performed.
Inspection is to identify the missing information RN, DAR, operator's signatures, shortages, etc.
2.
Control of Special Processes - An inspection was performed of the various processes used at ESD - Hunt Valley.
There were only three processes which required qualification of the process and/or personnel.
Soldering, Termi point, and wire wrap wiring were treated as special processes.
A sister division at the same site, Integrated Logistic Support Division (ILSD), performed the qualification of personnel except Termi point.
The qualifications of eight operators were inspected.
In this area nonconformances B.1 thru B.3 were identified.
In regard to noncon-formance B.1 it was noted that ESD - Hunt Valley had qualification procedures for soldering and Termi point, however, there was no procedure on wire wrap.
ILSD followed its own procedure to qualify the personnel.
In regard to nor anformance B.3 it was noted that the specification in question was added to the file during the inspection.
3.
Audits - The 1982 audit schedule was reviewed.
The audit plans and checklists used for three recent audits were inspected.
The reporting of results and follow up on findings were verified to follow established procedures.
The training of three auditors used for the above sample was inspected.
No nonconformances were identified in this area.
174
INDEX FACILITY REPORT NO.
PAGE NO.
L s
Anchor Darling Valve Company Williamsport, PA 99900053/82-01 1
Babcock and Wilcox Company Nuclear Power Generation Division Lynchburg, VA 99900400/82-01 5
Bechtel Power Corporation Gaithersburg Power Division Gaitherburg, MD 9990051 9/82-01 13 Bechtel Power Corporation Los Angeles Power Division No rwa l k, C A 99900521/82-01 19 Bechtel Power Corporation San Francisco Power Division San Francisco, CA 99900522/82-01 25 Burns and Roe, Incorporated Oradell, NJ 99900503/82-01 29 C&D Batteries Division of Eltra Corporation Plynouth Meeting, PA 99900765/82-01 35 Combustion Engineering, Incorporated Power Systems Group Windsor, CT 99900401/82-01 27 Comsip, Incorporated Customline Division Linden, NJ 99900771/82-01 43 Corner and Lada Company, Incorporated Cranston, RI 99900349/82-01 49 Dravo Corporation Pipe Fabrication Division Marietta, OH 99900017/82-01 53 Energy Incorporated Idaho Falls, ID 99900514/82-01 57 Exide Corporation Horsham, PA 99900359/82-01 61 The Foxboro Company Highland Plant East Bridgewater, MA 99900225/82-01 67 175
1 l
FACILITV REPORT NO.
PAGE NO.
l l
General Electric Company Wilmington Manufacturing Department i
Wilmington, NC 99900003/82-01 73 General Electric Company Nuclear Energy Business Operations San Jose, CA 99900403/82-01 77 Gibbs & Hill, Incnrporated New York, NY 99900524/82-01 87 ITT Grinnell Industrial Piping, Incorporated Kernersville, NC 99900019/82-01 91 ITT Grinnell Pipe Hanger Division, Engineering Department Providence, RI 99900285/82-01 97 Joseph Oat Corporation Camden, NJ 99900251/82-01 107 Limitorque Corporation Lynchburg, VA 99900100/82-02 111 Pacific Air Products Company Santa Ana, CA 99900769/82-01 119 The William Powell Company Cincinnati, OH 99900057/82-01 123 Siemens-Allis, Incorporated Small Motor Division Little Rock, AR 99900338/82-01 129 Square D Company Control Group Raleigh, NC 99900717/82-01 133 Stone and Webster Engineering Corporation Boston, MA 99900509/82-01 137 Teledyne Engineering Services Waltham, MA 99900513/82-01 145 Transamerica Delaval, Incorporated Engine and Compressor Division Oakland, CA 99900334/82-01 1 51 176
FACILITY REPORT NO.
PAGE NO.
United Engineers and Constructors, Incorporated Philadelphia, PA 99900510/82-02 1 61 Westinghouse Electric Corporation Electrical Systems Division Hunt Valley, MD 99900217/82-01 171 l
177 4
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NRC roav 335 1 REPORT NUV8E R (Aured by ODCJ U S NUCLE AR REGULATORY COMMISSloN
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BIBLIOGRAPHIC DATA SHEET NUREG-0040, Vol. 6, No. 2 4 T I T LE AN D SUB T I T LE (4 dd Votuma /Wo.,7 apperwresert 2 (Leave nim 14/
Licensee Contractor and Vendor Inspection Status Report Quarterly Report 3 RE CiPIENT'S ACCESSION NO April 1982 - June 1982
.9 PE HF OHVING OHGANi/A flON N AME AND MAILING ADORE SS flactu<a= /,p Code /
DATE REPORT ISSUED l U. S. Nuclear Regulatory Commission, Region IV vo3m Ivjaa982 Ju y 1 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 6 It * ** u- * /
8 Ileave Nanal 12 SPONSOHf NG ORG AN12 A flON N AVE AND M AILING ADDRE SS (incluaw I,p Codel p
11 CONTH ACT NO IJ T Y Pi OF HiPOHT PE RIGO C OVE RE D //nt/us we defeff 15 f,UPPi l ME N T A H V NOT( S 14 (L eave tw k /
, 16 AHS T H AC T (200 w ords o< 'cul This periodical covers the results of inspections performed by the NRC's Vendor Program Branch that have been distributed to the inspected organizations during the period from April 1932 through June 1982. Also included in this issue are the results of certain inspections performed prior to April 1982 that were not included in previous issues of NUREG-0040.
17 A E Y WOH OS AND DOCUYE N T AN AL YSIS l la DE SC H s P TOHS I 7t> IDE N TiF IE HS OPE N t N DE D TT AYS 18 AV AIL ARILIT Y ST ATE Yi N T 19 SE CU H t T V C L A SS ( 76,5 report /
21 NO OF P AG[ S Unclassified Unlimited u
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'7-t U.S. GOVERNMENT PRINTING OFFICE: 1982 361-302/39
UNITED STATES
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