ML20198K443
| ML20198K443 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/24/1986 |
| From: | Farrar D COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| 1155K, NUDOCS 8606040030 | |
| Download: ML20198K443 (3) | |
Text
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N Commonwealth Edison A
) One First National Plaza, Chicago, lihnois Address Reply to: Post Office Box 761
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Chicago. lilinois 60690 January 24, 1986 Mr. James G. Keppler Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137
Subject:
Byron Station Unit 2 IE Inspection Report No. 50-455/85-027 References (a): November 13, 1985 letter from J. M. Taylor to Cordell Reed (b):
December 12, 1985 letter from C. E. Norelius to Cordell Reed
Dear Mr. Keppler:
Reference (a) provided the results of the NRC Construction Appraisal Team (CAT) inspection conducted at Byron Station, Unit 2 on August 19-30, and September 9-20, 1985.
During this inspection, certain activities were found in violation of NRC requirements. Attachment A to this letter contains Cotunonwealth Edison's response to the Notice of Violation appended to reference (b).
Attachment B to this letter addresses the three construction program weaknesses identified in the CAT report. On January 8, 1986, Commonwealth Edison was granted a fourteen day extension on the due date for the response to the Notice of Violation.
With respect to certain findings identified during the CAT inspection, we do not believe these items represented a violation of NRC requirements. Our reasons for this are discussed in the detailed responses in Attachment A.
In these particular areas, we request the NRC to reconsider these items in light of the information we have provided.
Please direct any questions regarding this matter to this office.
i Very truly yours, l
l L. Farrar Director of Nuclear Licensing Im Attachments cc:
Byron Resident Inspector 1155K 8606040030 860124 h
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ATTACHMENT A VIOLATION la 10 CFR 50, Appendix B, Criterion III as implemented by Commonwealth Edison Company (CECO) Quality Assurance Manual (QAM), Quality Requirement No.
3.0, requires that measures shall be established to assure that applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, at the time of this inspection, the licensee's program was not adequately implemented in that splicing of Class 1E wiring in panels has occurred at Byron Station contrary to the FSAR commitments to IEEE Standard 420, which prohibits the use of wiring splices in panels. FSAR commitments had not been translated into appropriate procedures and design documents.
RESPONSE
This concern was originally identified during the NRC CAT inspection at Braidwood Station (IE Inspection Report No. 50-456/84-44; 50-457/84-40, page II-13, copy attached).
In response to this concern, a discussion of conformance to IEEE 420-1973, including our use of cable splices within control switchboseds, has been provided on page 8.1-14 of the FSAR.
An advanced copy of this FSAR revision was provided to the NRC in a February 6, 1985 letter from T. R. Tramm to H. R. Denton (copy attached).
During the Byron CAT inspection, a copy of this letter was furnished to the inspector.
Since our intent to formally revise the FSAR by way of amendment was committed to in the February 6, 1985 letter referenced above, we believe this item would be more appropriately classified as an Unresolved Item pending issuance of the formal FSAR amendment.
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9 The Braidwood Station FSAR commitment to IEEE Standard 420 prohibits the use of wire splices in Class IE equipment.
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However, NRC CAT inspectors observed in-line butt splices in numerous electrical panels.
As site procedures do not i
require the location of splices to be depicted on design documents, NRC CAT inspectors were unable to determine how extensively these splices have been utilized.
Additional-(
ly, the licensee had previously issued NCR-598 to document hardware deficiencies in installed butt splices and reported this condition to NRC Region III in accordance with 10 CFR 50.55(e).
The use of butt splices in Class 1E panels requires documentation in the FSAR as an exception to the IEEE standard.
The following are the isolated discrepancies noted by the NRC CAT inspector:
Conductor insulation damage on the orange conductor of cable IRH108-C1E in motor control center 1AP21E, cubicle F3.
ICR-7610 was subsequently issued to document this condition.
Several terminal screws were found loose in the Diesel Generator Control Panel 1A and in the Remote Shutdown Panel, section 1PLO5J.
ICR's 7646, 7644, and
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7643 were subsequently issued to document these condi-tions.
Internal motor lead T-9 was found damaged in motor operated valve 1CSQ01A.
ICR-7867 was subsequently
/
issued to document this condition.
/
The red conductor of cable ISIO53-C1E, in motor operated valve ISI8802A, was excessively bent and
/
not meeting minimum bend radius criteria.
ICR-7870 was subsequently issued to, document this condition.
(8) Seepage of 011 From Okonite Cable NRC CAT inspectors observed any oily substance seeping from jackets of numerous installed and terminated cables manufactured by the Okonite Company.
This condition was observed in both Class 1E and non-Class 1E cables in various Class IE equipment throughout the facility (motor control centers, main control boards, control panels, motor operated valves, etc.).
Information obtained from NRC Region III, CECO, and S&L revealed the following:
In a letter dated October 4,1982, including an attached engineering report (No. 364), the Okonite Company informed CECO that, with reference to the identical condition identified at Byron Station, this seepage "will not affect the reliability or life of the cables."
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- t C:mm:nwrith Edissa One First National P! ara Chicago. Illino.s I*
Address Reply to Post Off:ce Box 767 Chicago, tilinois 60690 February 6,1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Generating Station Units 1 and 2 Braidwood Generating Station Units 1 and 2 FSAR Changes NRC Docket Nos. 50-454/455 and 50-456/457
Dear Mr. Denton:
This letter provides advance copies of revised pages for the Byron /Braidwood FSAR. These changes are being made to provide more explicit descriptions of the design bases for these olants.
Enclosed is a revised page 8.1-14.
It now includes a discussion of conformance to IEEE 420-1973 with regard to the allowed use of cable splices within control switchboards. This change is being made to resolve a concern identified during a recent I&E inspection at Braidwood Station.
Also enclosed are revised pages E.20-1, E.20-la, and E.82-1 for Appendix E of the Byron /Braidwood FSAR. They now specify that the liquid
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source term used for shielding evaluation and environmental qualification of Byron and Braidwood does not include noble gases. Fission solids are the dominant contributor for long term doses so this does not significantly alter the radiological impact of postulated accidents.
Also included is a revised page E.21-3.
It specifies that backup sampling is not provided for hydrogen because of the large volume required.
l These changes will be incorporated into the FSAR at the earliest opportunity.
Please direct questions regarding these matters to this office.
One signed original and fifteen copies of this letter are provided l
for tRC review.
l Very truly yours, f
r e
-at. T. R. Tramm l
Wclear Licensing Administrator 1m
[QTh Y
cc: Byron Resident Inspector l
Braidwood Resident Inspector j
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.*e B/B-H /G APl.NDMI.5-38 MAY 1982
'7he physical identification of safety-related equipment is discussed in Subsection 8.3.1.3.
8.1.15 Shared Emergency and Shutdchn Electric systems for Multi-Unit Nuclear Power Plants The criteria followed in designing the two unit station is that each unit shall operate independently of the other and malf unc-tion of equipment or operator error in one unit will not initiate a malf unction or error in the other unit nor af fect the continued operation of the other unit.
8.1.16 Cuali fication of Class 1E Equiprent for Nuclear Pcwer Plants With regard to environmental qualification of instrumentation, control,. and electrical equipment important to safety, the Applicant complies with the intent of IEEE 323-1974 Additional inf ormation is provided in Section 3.11.
8.1.17 Availability of Electric Power Sources During ' abnormal electric power source configurations, pla nt operations are limited as described in Subsection 16.3/4.8.
8.1.18 Conformance to IEEE 338-1975 (Periodic Testing of Nuclear Power Generating Station Class 1E Power and Protection System) i conformance to this standard is addressed in Subsection 8.3.1.2 and 7.1. 2.19.
8.1.19 Conformance to IEEE 344-1971 (Fecommended Practices for
]
Seismic Oualification of Class 1E Equipment for Nuclear Power Generating Station)
Conformance to this Standard is addressed in Section 3.10.
8.1.20 Conformance to IEEE 387-1972 (Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating St at ions) i Vendor qualification tests, preoperational testing, and periodic testing during norral plant operation conform to those procedures i
described in this standard, except as ncted in subsections 8.3.1.2, 16.3/4.8, and Chapter 14.0.
t 8.1.21 Conformance to IEEE 420-1973 (IEEE Trial-Use Guide for Class lE Control Switchboards for Nuclear Power Generating Stations)
Class lE control switchboards conform to this standard with the following clarification to Paragraph 4.6.1.2:
Splices may be used on individual conductors of field cables within switchboards l
for the purpose of extending individual conductors to their point of termination.
8.1-14
I
/
l VIOLATION Ib 10 CFR 50, Appendix B, Criterion III as implemented by Commonwealth Edison Company (CECO) Quality Assurance Manual (QAM), Quality Rrquirement No.
3.0, requires that measures shall be established to assure that applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, at the time of this inspectlen, the licensee's program was not adequately implemented in that aphroximately one-third of the total A490 bolts tested by the NRC CAT were fodnd to be below the pretension required by AISC.
Installation and inspection requirements had not been translated into appropriate procedures for hiph strength bolted connections in structural steel and nuclear steam supply system joints which require pretension in the bolts.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED Structural Steel Connections There are sixty-five framing connections in the containment which have A490 bolts specified on design drawings. Twenty-nine of these connections are considered to be slip critical because the connection has slotted holes oriented parallel to the direction of the axial load.
These twenty-nine slip critical connections require some level of bolt pretensioning in order to transfer axial load.
The remaining thirty-six A490 connections have been qualified as bearint; type connections, and therefore, bolt torque need not be verified.
The slip critical connections were reinspecter! to determine the actual in-place torque value for each bolt in the connection. The connections were then evaluated by reducing the capacity of the bolts based on the ratio of as-found torque to the specified tocque. These reduced capacities were compared to design loads and the connection design was found to be within specified limits and the connections remain slip resistant. Theref ore, although the as-four.d pretension was below that specified under the applicable AISC provisions to provide the full load carrying capacity of the bolt, it was adecuate to meet the design.
Although the actual installed bolt conditions were evaluated and found to be acceptable, the bolts were brought up to specified pretension in order to restore margin.
Nuclear Steam Supply System (NSSS) Joint.s An engineering evaluation was performed to assess the adequacy of as-found bolt pretension in the NSSS supports. The engineering evaluation shows that the as-found bo!.t pretension is adequate to preclude separation of connections un:1er tension, and slip of friction type connections loaded in shear. Therefore, although the as-found pretension is below that specified under the applicable AISC provisions, it is adequate to meet the design.
i
I
. CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION Structual Steel Connections Although certain installation and inspection requirements were translated into the contractor's procedure, the following enhancement has been made to the procedure.
Blount Brothers Work Procedure Number 21 has been revised to clarify the installation requirements of high strength bolts by the turn-of-nut method. Snug tightening shall progress from the most rigid part of the connection to the free edges, and then bolts of the connection shall be retightened in a similar manner as necessary until all bolts are snugtight and the connection is compacted.
NSSS Joints Although certain installation and inspection requirements were translated into contractor's procedures, the following enhancements have been made to the procedure for installation of NSSS support bolts requiring pretension.
(1) Hunter Procedure SIP 4.001, " Bolted Connections", has been revised o
to define snugtight as 15% of the torque required to achieve the final pretension. This definition of snugtight will be used for future work on NSSS support high strength bolting.
(2) Connections which employ shims in the connected parts have been evaluated for their ability to achieve the required bolt pretension when tightened using turn-of-the-nut method. Where turn-of-the-nut method may not achieve the required pretension due to compressibi-lity of shims between the connected parts, the calibrated wrench method has been specified.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The engineering evaluations of the as-found condition of the structural steel connections and the NSSS joints were completed October 23, 1985.
The accessible bolts in the twenty-nine structural stoel connections were retightened by October 14, 1985.
Hunter Procedure SIP 4.001 was revised on October 15, 1985 and approved on January 9, 1986. Blount Brothers Work Procedure No. 21 was revised on November 13, 1985 and approved on January 5, 1986.
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VIOLATION Ic 10 CFR 50, Appendix B, Criterion III as implemented by Commonwealth Edison Company (CECO) Quality Assurance Manual (QAM), Quality Requirement No.
3.0, requires that measures shall be established to assure that applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, at the time of this inspection, the licensee's program was not adequately implemented in that during the inspection, concrete expansion anchors were found which did not meet the required bolt diameter embedment depth.
It could not be shown that embedment length requirements for concrete expansion anchors as specified in the concrete expansion anchor qualification report had been translated into appropriate installation and inspection procedures.
RESPONSE
The Standard Specification for Concrete Expansion Anchor Work, Form BY/BR/CEA, provides the installation and inspection requirements for concrete expansion anchors. Within this document is Figure 38-6, copy attached, which shows the embedded length (L ) measured from the surface of the concrete to the bottom of the expansion ring.
The qualification report for concrete expansion anchors is entitled, " Report on Static, Dynamic and Relaxation Testing of Expansion Anchors in Response to NRC IE Bulletin 79-02", dated July 20, 1981. We extracted two pages from Chapter III of this report, copy attached, which show the embedded length (leb) measured from the surface of the concrete to the bottom of the expansion ring. This consistently defines the embedded length.
The inspection to verify embedded length is a measurement of the projection of the anchor beyond the concrete in the installed position.
Subtracting this measurement from the total anchor length establishes the embedded length. The installed position is defined ao being after the anchor has been set by applying an installation torque. As the torque is applied, the anchor slips slightly, pushing the expansion ring outward to produce the wedging force between the ring and concrete.
During this setting action, it is assumed that the ring remains stationary as the back of the anchor approaches the expansion ring. The inspection for the embedded length is therefore, consistent with the qualification report.
Based on the foregoing, we believe embedment length requirements were properly translated into appropriate installation and inspection procedures. We are not aware of any concrete expansion anchors found during the CAT inspection that did not meet the required embedment length. We request the NRC to reconsider whether this is an example of violation of 10 CFR 50, Appendix B, Criterion III.
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SARGENT O LUNDY MtV. 2J ATTACHMENT 1 tNoiwttoo conc aco TABLE 38-2 MINIMUM EMBEDDED LENGTil, SPACING AND EDGE DISTANCE FOR EXPANSION ANCH0PS' Nominal Minimum Embedded F.inimum Spacing.
Minimum Edge Distance Bolt length *
(inches)
(inches)
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(S)
(inch)
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(ES) 1/4 5/8 2.5 3.25 1.75 3/8 3
4.5 5
2.5 1/2 4
6 7
3.5 5/8 5
7.5 8.5 4.75 3/4 6
9 10 5
1 8
12 13 6.5
- Minus 1/16" tolerance is allowed.
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CIIAPTER III - MATERIALS AND TEST SPECIME!!S EXPA! SIO:: A?:CHOMS J[escription of Generic Types.
In this expericental program, The four generic types of expansion cnchors have been investigated.
generic types are classified into the following catagories: wedge, siceve, self-drilling and drop-in.
All the generic types listed above achieve lead carrying capacity when embedded by having a wedge ecchanisa located at the bottom of the anchor.
The wedge is expanded against the side walls of the embede2nt hole during the installation procedure.
Figure 3.1 shows the four generic types cf cnchors that were tested and also identifies the reanufacturer of the specific type investigated.
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For each anchor, the ethedment depth has consistently been i;s defined as the distance from the surface of the erbcdding material to g'l a
t' 'l the bottoa of the expanding part.
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This aucunes that the anchor is G
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VIOLATION 2 i
10 CFR 50, Appendix B, Criterion VII, as implemented by CECO QAM, Quality Requirement No. 7.D. requires measures shall be established to assure that purchased material, equipment, and services conform to the procurement documents.
Contrary to the above, at the time of this inspection, the NRC CAT inspectors found several deficiencies in vendor supplied components. The deficiencies included:
radiographic film stored by the component supplier in an off-site facility were not retrievable.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED This violation involves the radiographs for the Unit 2 component cooling surge tank and volume control tank. The component cooling surge tank was fabricated, inspected, tested and shipped from the Westinghouse Orange Plant. Radiography was required and performed in accordance with ASME Section III Class 3 requirements.
This product line was transferred to other Westinghouse facilities and Westinghouse considers the radiographic films for this tank to be lost in transit between facilities.
Although the actual radiographs are not available, the original radiograph procedures, shooting sketches, and reader sheets which document the performance and acceptance of the radiography are available. The reader sheets are permanent records which can be copied and do not degrade with time.
With respect to radiographs for the volume control tank, Westinghouse has confirmed that these films are present in their respective storage locations.
Since the radiographs for the component cooling surge tank are not retrievable, the Westinghouse Product Assurance Department conducted an inventory of the radiographs for Byron /Braidwood equipment for which they are responsibic. The results of the inventory indicate that all radiographs that are required by Westinghouse quality release are either in the Westinghouse storage facility or in their vendors' storage facilities with the exception of the component cooling surge tank.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION Commonwealth Edison's Quality Assurance Department has added radiograph retrievability as a quality element in their next scheduled audit of Westinghouse.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The Westinghouse inventory was completed October 18, 1985.
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VIOLATION 2 (Cont'd) 10 CFR 50, Appendix B, Criterion VII, as implemented by CECO QAM, Quality Requirement No. 7.0, requires measures shall be established to assure that purchased material, equipment, and services conform to the procurement documents.
Contrary to the above, at the time of this inspection, the NRC CAT inspectors found several deficiencies in vendor supplied components. The deficiencies included: undersized welds were identified on tanks and heat exchangers.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED Prior to the CAT inspection, undersized welds were identified by Commonwealth Edison on ten tanks and pressure vessels as a result of a walkdown performed in response to NRC Information Notice 85-33.
Three Commonwealth Edison nonconformance reports (NCR's) were issued to track resolution of these deficiencies.
These NCR's were evaluated by the Project Engineering Department and the equipment vendors and the as-found condition of the components was found acceptable.
During the CAT inspection, NRC inspectors identified additional tanks and pressure vessels with undersized welds. Hunter Corp. NR 1148 was issued to address these items and is currently under evaluation by Sargent &
Lundy.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION Although the undersized welds which have been identified thus far have been found to not be design significant, approximately 20 other potentially affected tanks and pressure vessels have been inspected and deficiencies which were identified have been included into the scope of NR 1148. The welds in approximately six additional potentially affected tanks are not optimally accessible at this stage of construction.
If the results of the engineering evaluation of the other components in NR 1148 indicate that there are no undersized welds of design significance, we do not believe any further inspection effort is warranted.
DATE WHEN FULL CONpLIANCE WILL BE ACHIEVED The engineering disposition of Hunter Corp. NR 1148 is expected to be complete by June 30, 1986.
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VIOLATION 2 (Cont'd) 10 CFR 50, Appendix B, Criterion VII, as implemented by CECO QAM, Quality Requirement No. 7.0, requires measures shall be established to assure
+ hat purchased material, equipment, and services conform to the procurement documents.
Contrary to the above, at the time of this inspection, the NRC CAT inspectors found several deficiencies in vendor supplied components. The deficiencies included:
various vendor radiographs did not have complete weld coverage or did not show the required weld quality.
This response addresses the eight notes of Table IV-6 discussed on pages IV-21 and IV-22 of the CAT Inspection Report.
Note 1 l
1 The film packet for one of the welds which should have contained film for five intervals, contained film for only two intervals. The missing film for the other three intervals was later found and was reviewed with no problems identified.
RESPONSE
The film packet noted was previously transferred to the station vault 4
from the Quality Assurance Department and during the station's indexing and boxing of the film, the film was divided and separated. This separation of documents was corrected during the CAT inspection and the i
complete set of film was reviewed and found acceptable by the CAT inspector. We believe we have demonstrated the required documents are retrievable and this item should not be considered as an example of violation of 10 CFR 50, Appendix B, Criterion VII.
Note 2 The reader sheets with the film indicated that this film was for Braidwood components.
Further investigation of documentation indicated that the items had originally been scheduled for Braidwood but had later been transferred to Byron.
RESPONSE
procurement specifications for Byron and Braidwood Stations are common documents. Materials and components which are procured under these common specifications are usable at either station. This component was originally planned for installation at Braidwood Station, but was subsequently transferred to Byron and installed. Transferring components from one site to the other is a common, acceptable practice. The transfer of this component from Braidwood to Byron was properly documented by Sargent & Lundy prior to the CAT inspection. A review of the reader sheets and associated film for this component indicated the weld's identification numbers and weld quality were acceptable. Therefore, we i
do not consider this item to be an example of violation of 10 CFR 50, Appendix B Criterion VII and request the NRC to reconsider their disposition of this item.
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Reader sheets for weld WT 8549-64 were not in that packet but were found in film packet for weld WT 2158-64.
RESPONSE
^
A review of this item indicated that the reader sheet was inadvertently placed in the different packet (box). The radiographs are indexed by Sargent & Lundy transmittal numbers which, generally, list several different welds. Occasionally, it is not physically feasible to store all the radiographs and associated documentation in the same box.
Therefore, when the CAT inspector requested a random sample of film associated with certain specifications, one box was provided. To preserve the randomness of the sample, the box selected was not reviewed to determine if the reader sheet was contained within the box prior to being presented to the CAT inspector.
If the box would have been screened by plant personnel prior to being presented to the inspector, the misplaced reader sheet would have been included with the box selected. Therefore, we believe this item should not be considered an example of violation of 10 CFR 50, Appendix B, Criterion VII and request the NRC to reconsider their disposition of this item.
Note 4 (1st Paragraph)
Circumferential seam 1Al-6 showed added veld metal on interval A-B dated June 20, 1977, however the added metal does not show in interval M-A dated June 23, 1977. Ceco reviewed the film and returned it supposedly in proper order. However, due to two different identifications on the l-film, it was impossible to tell whether on\\ or two welds were represented i
by these film.
Subsequently,CEcodetermirgdthatthefilmactually were from only one weld, and that further rtpair in the M-A interval j
accounted for the difference in appearance c the two weld intervals.
RESPONSE
i During the radiographic review, it appeared tc the CAT inspector that l
weld metal was added subsequent to the initial acceptance of the weld.
l However, the proper indexing of the film package by plant personnel clearly demonstrated the sequence of repairs an resolved the inspector's concern. We do not consider this item to be an example of violation of 10 CFR 50, Appendix B, Criterion VII and request the NRC to reconsider their disposition of this item.
i Note 4 (2nd Paragrapa)
On weld 1Al-1 (Job 2) interval 11-12 a linear indication, possibly a i
crack, was noted.
Ceco then did an ultrasonic examination of the area.
However, the performed examination was a longitudinal examination and is j
not acceptable for this type and location of indication. An NCR is being written. This indication is in vessel ICS01T (Byron 1).
. I CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED This tank is a Unit 1 component and therefore, Action Item Record
- 6-85-351 was written to track resolution of this issue. The weld was re-radiographed, surface buffed and again re-radiographed. The indication was a surface condition, not a crack, which was removed during the buffing.
The aforementioned repair film was subsequently reviewed by NRC Region III Inspector K. D. Ward and found acceptable.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION We believe the presence or absence of this indication was a matter of judgment by Level III interpreters. As a conservative measure, the weld was re-radiographed and the indication was removed. Based on the large number of radiograph films that were reviewed by the CAT and found acceptable (approximately 1980), we do not believe any further action is warranted.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The second re-radiograph, which showed the surface condition was removed, was completed on November 5, 1985.
Note 4 (3rd Paragraph)
In vessel H1 tank 2 weld 3Al-1 interval F-G at the end of the seam in the transverse weld, an indication was noted. However, when the film of the transverse weld were located they showed no indication of a problem in the area mentioned.
RESPONSE
The radiographing of the transverse weld was performed subsequent to the seam weld radiograph and no indication was noted in the transverse weld. Tht: Indication was considered a surface condition and was removed prior to radiographing the transverse weld. This is an acceptable practice and we do not consider this item to be an example of violation of 10 CFR 53, Appendix B, Criterion VII.
Therefore, we request the NRC to reconsider their disposition of this item.
Note 5 When the film was submitted, the reader sheets were so faint that it was impossible to read them. Subsequently, the original sheets in the QA file were produced and the film was read. No problems were noted.
. RESPONSE The onsite copy of the reader sheet for this film was reproduced from a document which was faintly printed. A better copy was made from the original reader sheets used for tha initial review and acceptance by Sargent & Lundy.
This legible copy was reviewed by the CAT inspector and found acceptable. We do not consider this item to be an example of violation of 10 CFR 50, Appendix B, Criterion VII and therefore request the NRC to reconsider their disposition of this item.
Note 6 A 3/8 inch slag line was noted in the 2-3 interval. After reviewing the receiving and issuing documents, it was determined that the item had never been issued to the job for installation.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED This component was a Unit 1 pipe hanger (M-1RCl20115) which was eliminated from the piping design.
Although this hanger was originally accepted for use, elimination of it from the design will prevent it from being installed.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATICN This hanger was the only component supplied by ITT Grinnell which required a radiographic exam. Therefore, we do not believe any further action is warranted.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED This component was eliminated from the design on February 15, 1980.
Note 7 Area 3255-A OWX10D had a linear indication on each of the junction welds. Area 3244-A-2V2 OWXO7 TV2 had the penetrameter shim into the area of interest.
RESPONSE
These two components are non-safety related tanks.
Furthermore, we believe these concerns resulted from a subjective judgment involving interpretation of radiographs.
These radiographs were originally reviewed by a Sargent & Lundy Level III interpreter and found acceptable.
In response to this concern identified during the CAT inspection, the
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. radiographs were re-evaluated by Sargent & Lundy's Level III interpreter. With respect to tank OWK10D, the linear indication is considered to be a 1/4" long weld surface contour and is acceptable.
The penetrameter shim in tank OWXO7 is along the toe of the weld and does not impair the diagnostic capabilities of the radiograph. This is also considered acceptable. Therefore, we do not consider these items to be an example of violation of 10 CFR 50, Appendix B, Criterion VII and request the NRC to reconsider their disposition of these items.
Note 8 4901-9, weld 147 A-B, had an unacceptable slag line at Station A and the belt numbers appear to be inside the area of interest.
Six welds did not have full coverage. The welds were identified as:
4901-9, - weld 103 G-H, weld 116 A-B, weld 160 G-H; 4901 welds 147 A-B and 160 G-H; and 4902, weld 75 B-C.
Item 4901-1D, weld 103 G-H, also did not have full coverage and there was a linear indication extending into the uncovered area.
RESPONSE
With the aid of additional drawings incorporated into an approved Engineering Change Notice (ECN) that was not available at the time of the CAT inspection, it can be demonstrated that the slag line and belt numbers discussed above are located in the base metal. The slag inclusion was actually a surface blemish caused during the final surface preparation. This additional information was used by Sargent & Lundy during their initial review of the radiographs and was found acceptable.
The additional drawings also demonstrate that the radiographs obtained the required coverage of the area of interest. The ECN and radiographs were subsequently reviewed by NRC Region III Inspector K. D. Ward and found acceptable. Therefore, we do not consider these concerns to be an example of violation of 10 CFR 50, Appendix B, Criterion VII and request the NRC to reconsider their disposition of these items.
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N VIOLATION 2 (Cont'd) 10 CFR 50, Appendix B, Criterion VII, as implemented by CECO QAM, Quality Requirement No. 7.0, requires measures shall be established to assure that purchased material, equipment, and services conform to the procurement documents.
Contrary to the above, at the time of this inspection, the NRC CAT inspectors f'ound several deficiencies in vendor supplied components. The deficiencies included:
fasteners for various components (large pump-motor assemblies, battery racks, switchgear, other electrical equipment, and HVAC equipment) were not of the material required by specifications or drawings.
CORRECTIVE ACTI9N TAKEN AND RESULTS ACHIEVED Large Pump-Motor Assemblies Commonwealth 3dison non-conformance report (NCR) F-1014 was issued to track resolution of the discrepancies regarding assembly and mounting bolts for large vendor supplied pump-motor assemblies. Sargent & Lundy
-and Westinghouse have reviewed the NCR and evaluated the installations where the as-found bolting either differed from specifications in design documents or could not be shown to meet the design specifications. Their conclusions are that the as-found bolting meets or exceeds the strength requirements intended by the design specifications, except for the motor-to-base bolts on the non-safety related positive displacement charging pump (2CV02P).
Further analysis is necessary to determine the acceptability of the motor end-to-pump base bolts on the RHR pumps (2RH0lPA and 2RH01PB) and the motor-to-skid bolts on centrifugal charging pump 2CV01PB.
The motor-to-base bolts on pump 2CV02P will be replaced with bolts which can be shown to meet design specifications.
If the analyses of the motor end-to-pump base bolts on pumps 2RH0lPA and 2RH01PB and the motor-to-skid bolts on pump 2CV01PB do not yield acceptable results, these bolts will also be replaced.
ASTM A307 Bolts for Electrical and HVAC Equipment This issue concerns the installation of unmarked bolts in various components whose design requirements call for ASTM A307 bolts. The issue has been categorized into two subgroups:
(1) unmarked bolts which were site procured and installed by site contractors; and (2) unmarked bolts provided with vendor supplied equipment. Commonwealth Edison issued NCR's F-1001 and F-1012 to address these subgroups, respectively.
As a result of these NCR's, a sampling plan has been devised to perform a Brinnell hardness test on unmarked bolts in a random selection of equipment in both subgroups. The correlation of Brinnell hardness to tensile strength will be used to establish the acceptance basis for the unmarked bolts.
All samples have been taken and the data has been evaluated by Sargent &
Lundy. The unmarked bolts have been found acceptable.
. CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION Large Pump-Motor Assemblies In October 1982, the Commonwealth Edison Co. Quality Assurance Manual was revised to require the random inspection of bolting materials in all fabricated assemblies upon receipt of the equipc.ent to assure compliance with design drawings. The equipment included in NCR F-1014 was received prior to 1982.
ASTM A307 Bolts for Electrical and HVAC Equipment On August 19, 1985, the Byron Project Construction Superintendent issued a letter to the applicable onsite contractors concerning NCR F-1001. This letter directed the contractors to perform a sample inspection during receipt of future shipments of bolts to verify the bolts are marked per ASTM A307 requirements.
The revision to the Quality Assurance Manual discussed above also addresses vendor supplied electrical and HVAC equipment similar to the equipment included in NCR F-1012.
The equipment included in NCR F-1012 was received prior to 1982.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Large Pump-Motor Assemblies
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The motor-to-base bolts on pump 2CV02P will be repl' aced by May 1, 1986.
The analysis of the motor end-to-pump base bolts on pumps 2RH01PA and 2RH0lPB and the motor-to-skid bolts on pump 2CV01PB will be completed by March 1, 1986.
ASTM A307 Bolts for Elec WJJ')_a d MVAC Equipment g
Sargent & Lundy's eval r.t sn t,
- he Brinnell hardness data from' bbe sample plan was completed on January 22, 1986.
Aa a result of Sargent &
Lundy'c evaluation, the aforementioned unmarked bolts were determined to be neceptable.
Commonwealth Edison NCR's F-1001 and F1012 were closed on Janesey 24, 1086.
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3 VIOLATION 3a 4
' 10 CFR 50, Appendix B, Criterion X, as implemented by CECO QAM, Quality Requirement 10.0, requires that. a program for inspection of rptivities shall be established and executed to verify conformance with documented instructions, procedures, and drawings for accomplishing the s
activities.
4 Contrary to the above, at the time of this inspection, the licensee's inspection programs were not effectively implemented in that Unit 2 4160V switchgead and DC fuse panels were found not to be installed in accordance with requirements for seismic mounting of Class IE equipment.
l CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED 4160V Switchmear Commonwealth Edison nonconformance report (NCR) F-1005 was issued to document the nonconforming hold down welds on switchgear 2Ap05E and 2Ap06E.
In addition, two field change requests (FCR's) were written. FCR F-26662 was written to allow revision of the hold down weld design from a four-y sided weld to a two-sided weld. FCR F-26659 was written to allow an alternate hold down weld design where inspection was deterred due to the location of the two slots in the rear of the cubicles. These FCR's were
- evaluated by Sargent & Lundy and approved.
The no.nconforming hold down welds were repaired and reinspected.
DC Fuse panel Hatfield Electric Co. Discrepancy Report No. 7373 was issued to address the' mounting weld configuration for DC fuse panel 2DC11J. The mounting welds were repaired and reworked to meet the weld configuration require-ments of the specification. This rework was reinspected and found acceptable.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION As a result of a stop work order regarding installation of electrical equipment that was issued in December, 1980, the installation and inspection practices in this area were significantly enhanced and became more i
prescriptive and rigorous. The switchgear and fuse panel discussed in this violation were installed prior to this overall upgrade to the electrical equipment installation process. Therefore, we believe current procedures and practices should be sufficient to prevent recurrence.
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. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED NCR F-1005 concerning the hold down welds on switchgcar 2AP05E and 2AP06E was closed on December 14, 1985.
Discrepancy Report No. 7373 concerning the mounting weld configuration for DC fuse panel 2DC11J was closed on September 9, 1985.
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i, j-VIOLATION 3b i
10 CFR 50, Appendix B, Criterion X, as implemented by CECO QAM, 4
Quality Requirement 10.0, requires that a program for inspection of 1
activities shall be established and executed to verify conformance with documented instructions, procedures, and drawings for accomplishing the I
activities.
Contrary to the above, at the time of this inspection, the licensee's inspection programs were not effectively implemented in that some Class lE j
electrical raceways have not been installed in accordance with FSAR commitments for electrical separation.
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RESPONSE
l Commonwealth Edison commitments with regard to electrical separation are documented in Section 8.3 of the ByroniBraidwood FSAR. These commitments are supplemented with a description of conformance to NRC Regulatory Guide i
1.75 which is documented in Appendix A of the FSAR.
Byron's commitment to Regulatory Cuide 1.75 allows for the use of analysis and/or test to justify l
separation distances less than those distances specifically stated in Section i
8.3 of the FSAR.
The use of analysis and/or test to justify electrical separation dictances is also permitted in IEEE Standard 384-1974.
J Prior to this CAT inspection, a Sargent & Lundy analysis justifying
" worst case" separation distances between safety related and non-safety j
related cables was submitted to NRR for review. The specific configuration which was chosen for a " worst case" analysis was one in which a separation 4
distance was established between a safety-related and a non-safety related cable when one is in free air and the other is in a raceway. The analysis l
justified that a separation distance of less than one inch is acceptable between the cable and the raceway. Thisanalysiswasbasedonagest performed for Byron Station.
i The specific raceways which were identified by the CAT as being in l
violation of the FSAR commitments involved installations where either non-safety related conduits were installed with less than 12" vertical or 3" horizontal separation from safety related cable tray, or safety related conduits were installed with less than 12" vertical or 3" horizontal separation from non-safety related cable tray.
In order to satisfy the CAT inspector's concern for the type of installations identified, the CAT was presented with Sargent & Lundy Calculation 4391/Q-15, " Justification of Electrical Separation Distance Between Safety Related and Non-Safety Related Raceways". This calculation is based on the same test report which was submitted to NRR for review and justifies that separation of less than 12" vertical or 3" horizontal, but greater than one inch, is acceptable between a safety related conduit and l
non-safety related cable tray or between a non-safety related conduit and a safety related cable tray, t
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1 Since Sargent & Lundy Calculation 4391/Q-15, as well as the calculation previously submitted to NRR, justifies separation of greater than one inch between safety related and non-safety related cable tray and conduit, the only specific electrical separation inspection requirement which is required to be in the electrical contractor's inspection procedures is the requirement to verify that one inch separation is maintained. Hatfield Electric Co. Quality Control procedure 9B has a requirement that specifies one inch separation between cable tray and conduit installations.
Based on the information provided above, we believe this item would be more appropriately classified as an Unresolved Item pending the outcome of NRR's review of the Sargent & Lundy analysis.
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VIOLATION 3e 10 CFR 50, Appendix B, Criterion X, as implemented by CECO QAM, Quality Requirement 10.0, requires that a program for inspection of activities shall be established and executed to verify conformance with documented instructions, procedures, and drawings for accomplishing the activities.
Contrary to the above, at the time of this inspection, the licensee's inspection programs were not effectively implemented in that some Class IE motor operated valve terminations were not accomplished in accordance with design documents in that wiring configurations did not match those specified on approved wiring diagrams.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED As a result of the examples of wiring discrepancies identified during the CAT inspection, the electrical installation contractor initiated nonconformance report (NCR) 1697 to address this concern. Twelve valves were reinspected under this NCR.
The as-wired condition in each valve was examined and it was determined that the valves would perform their design function.
This is because each valve was properly wired in accordance with the control schematic diagram which is the governing design document. We acknowledge that in four valves, the wiring was not in accordance with the wiring diagram for the valve. Since a wiring diagram is also a design document and is used for QC inspection purposes, the affected wiring diagrams were revised to reflect the as-wired condition of the four valves.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATION Training sessions will be conducted for onsite electrical contractor inspection personnel to re-emphasize the procedural requirements for documenting the installation of field changes that affect design documents issued for construction.
In addition, training sessions will also be conducted for appropriate onsite Commonwealth Edison personnel to re-emphasize the established guidelines and methods for installing and documenting field changes that affect wiring diagrams issued for construction.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Revisions to the affected wiring diagrams were initiated on October 21, 1985. The training sessions discussed above are expected to be completed by February 21, 1986.
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s ATTACHMENT B CONSTRUCTION PROCRAM WEAKNESS 1 For two samples of radiographs for ASME components supplied by Westinghouse (W) and stored in (W) facilities which were requested by the NRC CAT for review, none were provided. This is indicative of a lack of retriev-ability for ASME Code required documentation and raises questions whether code documentation is available for the (W) supplied equipment.
In addition, the NRC CAT review of audits by CECO and (W) indicated that audits had not addressed the area of retrievability of radiographs.
RESPONSE
In response to this concern, the Commonwealth Edison Manager of Projects and the Westinghouse Manager of Commonwealth Edison Projects became involved. The Westinghouse Product Assurance Department was directed to conduct an inventory of radiographs for all Byron and Braidwood equipment for
. which Westinghouse is responsible. The results of the inventory indicated that all radiographs that are required by Westinghouse quality release are either in the Westinghouse storage facility or in their vendors' storage facilities with the exception of the component cooling surge tank.
Westinghouse considers these radiographs to be lost, however other documents related to these radiographs are available which support the acceptance of the radiography on this tank.
In addition, the Commonwealth Edison Quality Assurance Department has been directed to add radiograph retrievability as a quality element in their next scheduled audit of Westinghouse.
They will check Westinghouse's audit activity in this area as well as their ability to retrieve radiographs.
We believe the appropriate level of management attention has been given to this issue and the actions taken address not only the Byron Project, but also the Braidwood Project.
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CONSTRUCTION PROGRAM WEAKNESS 2 A significant number of A490 bolts used in structural steel connections and equipment hold-down applications were found by the NRC CAT to have less than specified torque values. Some of these connections are designed to rely on bolt induced clamping forces to carry a portion of the expected loads.
RESPONSE
As this issue was identified during the course of the CAT inspection, an engineering evaluation of the problem was performed. The evaluation considered the as-found condition of the affected bolted connections in Unit 2 structural steel and NSSS supports. The results of this evaluation indicated that the bolted connections found with torque values below the inspection torque had no design significance.
Since Byron Unit 1 was operating at the time of this inspection, this concern was further evaluated to determine any potential effect on Unit 1 operation. This was accomplished by applying the as-found torque values of the bolts in the affected Unit 2 connections to the corresponding bolted connection in Unit 1.
From this evaluation, it was concluded that there should not be any hardware deficiencies of design significance in Unit 1.
In order to further support this conclusion, a commitment was made to reinspect the corresponding bolted connections on Unit 1 during the next scheduled outage.
This was accomplished during the October - December, 1985 outage.
The as-found torque values in the corresponding Unit I structural steel bolted connections were similar to those found in Unit 2.
The detailed engineering evaluation of these Unit 1 connections likewise concluded there was no design significance in the as-found condition. However, the bolts found to have torque values less than the inspection value were retensioned to restore margin.
The as-found torque values in the corresponding Unit 1 NSSS support steel connections were similar to those found in Unit 2, except some of the Unit 1 bolts exhibited lower as-found torque values than the corresponding connections in Unit 2.
Therefore, all Unit 1 connections of this type were reinspected. A detailed engineering evaluation of the as-found condition of all these connections concluded there was no design significance. Nevertheless, all the bolts found to have torque values less than the inspection value were retensioned to restore margin.
Based on the discussion above for Unit 1 and the response to Violation Ib concerning Unit 2, we believe the appropriate level of management attention has been applied to this issue to assure that completed installations meet design requirements for bolted connections in structural steel and NSSS supports.
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r CONSTRUCTION PROGRAM WEAKNESS 3 Examples were found in which the electric wiring for motor operated valves were not in accordance with approved design drawings. This is of further concern in that the methoj used for QC to accept these installations was through the use of a " speed memo" (which is an uncontrolled document that does not receive the appropriate design review and approvals). Also in the electrical area, the foundation mounting welds of several pieces of Class IE 4160V switchgear and 125V DC fuse panels were not in accordance with design requirements.
RESPONSE
Wiring for Motor Operated Valves A number of motor operated valves identified by CAT inspectors contained wiring which was not terminated at the termination points identified on some design drawings. The as-found wiring terminations resulted in electrical circuitry which was functionally acceptable when compared to the control schematic diagram. However, the specific details of some termination points were not per the wiring diagram. These minor discre-pancies were corrected by revising the wiring diagrams to reflect the as-wired condition.
During the CAT inspection, an evaluation of the as-found Unit 2 wiring discrepancies on the corresponding Unit 1 valves resulted in the same conclusion. The valves would perform their design function.
During the October - December, 1985 Unit 1 outage, the affected valves in Unit 1 were reinspected to confirm this.
It was determined that all the valves' wiring was terminated per the wiring diagram, as well as per the control schematic diagram.
The " speed memo" discussed above only provides wiring details which supplement design details shown on wiring diagrams. This memo does not provide any instructions or details which would result in a valve being wired such that it would not conform to approved wiring diagrams.
Consequently, there is no requirement to control this memo as a design document.
Based on the preceeding discussion, we have concluded that the completed installations of motor operated valves meet design requirements and no further management attention is warranted.
Mounting of Electrical panels and Switchmear As identified in the CAT Inspection Report, deficiencies were observed by NRC inspectors in hold down welds associated with certain switchgear units and a fuse panel. The report further stated that similar deficien-cies were previously identified by the licensee on Unit 1 equipment.
These similar Unit I deficiencies had been identified as a result of the Quality Control Inspector Reinspection program which was a corrective
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. Mounting of Electrical Panels and Switchgear (Cont'd) action program executed as a result of NRC Violation 50-454/82-05-19.
That program was executed to verify that inspections performed by Quality Control Inspectors prior to September, 1982 were valid inspections. The discrepancies found upon reinspection were evaluated and found to have no design significance.
It was concluded that no further expansion of the reinspection effort was necessary. Further to this activity, as a result of the corrective actions taken in response to NRC Violation 50-454/80-25, the electrical installation procedures were revised in the 1981 time frame to be more prescriptive and detailed. As a result of the foregoing, we judge that no additional management action is warranted with regard to electrical equipment mounting and that completed installations in this area meet design requirements, ll55K