ML20195F699
| ML20195F699 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/20/1984 |
| From: | BROWN & ROOT, INC. (SUBS. OF HALLIBURTON CO.) |
| To: | |
| Shared Package | |
| ML20195F283 | List:
|
| References | |
| FOIA-85-59 CP-CPM-6.9B, NUDOCS 8606120264 | |
| Download: ML20195F699 (32) | |
Text
{{#Wiki_filter:. - _ - _ _ y PROCEDURE EffECTlyE BROWN "* DATE PAGE NUMBER REVISION CS i _.. . , JOB 35-119e a CP-CPM 6.9B 2 09/21/84 1 o f 13 TITLE: ORIGINATOR: / 6 b8-Y U OATE (APPENDIX B) REVIEWED BY: f 1/ 69 A R C NTROL REVIEWED BY: .dNg/n/w /((M A c,41 77b ' TUGC0 QUALITY ASSURANCE DATE' REVIEWED BY: J/[ M h> 9M'o/N APPROVED BY: 9 /f# CONSTRUCTION PROJECT MANAGER ' ~0 ATE / o.1 TABE OF CONTENTS
1.0 INTRODUCTION
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g 2.0 GENERAL / 2.1 APPROVAL AUTHORITY 1
2.2 RESPONSIBILITY 3.0 WELD FILER MATERIAL CONTROL PROCEDURE j 3.1 PROCUREMEhI 3.2 STORACE OF Q WELD FILLER MATERIAL 3.3 ISSUANG OF WELD FILER MATERIAL 3.4 IDENIIFICATION OF INCONSISTANCIES 3.5 DISPOSITION OF RETURNED WELD FILIER MATERIAL 3.6 DISPOSITION / ISSUE /USE OF NONCONFORMING WELD FILLER MATERIAL o.11 TABES 6.9 B-I WELD FILER MATERIAL PROCUREMENT STANDARDS INDEX o.iii FIGURES 6.9 B-1 OPERATION LOG FOR STATIONARY AND PORTABE ROD OVENS 6.9 B-2 WELD FILLER MATERIAL LOG 6.9 B-3 WQTC FILLER METAL USE LOG l
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o .1v SUPP LEMENT 6.9 B-I PROCUREMENT OF WELD FILLER MATERIAL 8606120264 860512 PDR FOIA GARDE 85-59 PDR
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BROWN & ROCI, INC. PROCEDURE EFFECTIVE
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' ' CPSES NUMBER REVISION DATE PAGE r.
JOB 35-1195 a CP-CPM 6.9B 2 09/21/84 2 o f 13 1.0 INTRODUCIION l 4 This appendix delineates requirements for the procurement and ! control of all welding filler material. 2.0 GENERAL 2.1 APPROVAL AUT.HORITY The requirements for origination, review, and approval of this appendix shall be in accordance with procedure CPM 6.1. In addi-
;- tion, the appendix and its' DCN's shall be by the Project Welding Engineer.
2.2 RESPONSIBILITY Compliance with this appendix shall be the responsibility of the applicable Craft General Superintendant (CGS) and the Project Welding Engineer (PWE).
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3.0 WELDING FILER MATERIAL CONTROL PROCEDURE 3.1 PROCUREMENT Procurement of r.id filler material shall be in accordance with So,,plement 6.9 e-I. 3.2 STORACE OF Q WELD FILER MATERIAL i 3.2.1 Material Distribution Station and Main Storage "Q" weld fi(ler naterial shall be stored in a level B facility in accordance with MCP-10. The temperature of the storage facility shall not be below 40*F.
"Q" weld filler m.iterial original containers shall, upon receipt and during storage, be clearly marked with the weld filler material classification, size, and heat / lot number.
Receipt and transport of all weld filler material from main storage shall be controlled. During storage in nain storage facilities, "Q" weld filler material shall remain in the original containers. J 'y-I w.- W .. _ . . _ . . . _ . _ . _ . _ .
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BROWN & ROOr, INC. CPSES PROCEDURE NUMBER REVISION DATE PAGE ! JOB 35-1195 CP-CPM 6.9B 2 09/21/84 3 o f 13 Access to the Material Distribution Station (MDS) shall be controlled by the GMS and limited to Welding Engineering, Quality Control, the Authorized Nuclear Inspector and the authorized Pipe Department personnel. All associated warehouse requisitions for weld materials shall be maintained at the MDS until the material is used or removed from the MDS.
"Q" weld filler material shall remain in original containers during transport and prior to issuance except low hydrogen electrodes which may be issued from stationary ovens or from their containers. Unissued electrodes shall be placed immediately in a heated stationary oven or portable container.
3.2.2 Rod Ovens Heated stationary or portable rod ovens shall be checked by MDS
. personnel as follows:
? 1. Checking Frequency. Stationary rod oven temperature shall be checked before first used or daily while in use. Portable ri i oven temperature shall be checked monthly or whenever t'aey are suspected of malfunction and prior to reuse after repair. Portable heated rod containers are , ? checked for operation before being issued to craft personnel. ' 2. Records. A log of stationary and portable rod oven temperatures as ' required above shall be maintained by construction personnel (Figure 6.9B-1) . Such records will be available for inspec-t io n. Completed log sheets shall be forwarded to the Permanent Plant
- Records Vault.
The PWE shall perform a surveillance of the MDS every two weeks as a minista and doctanent the surveillance on a Surveil-lance Checklist. e r* %
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~T BROWN & ROCI, INC. PROCEDURE NUMBER REVISICN DATE PAGE CPSES JOB 35-1195 CP-CPM 6.9B 2 09/21/84 4 o f 13
- 3. Checking Rod Ovens.
The procedure for checking stationary or portable heated rod ovens is as follows:
- a. Activate oven.
- b. Allow 60 minutes for oven to warm up.
EXCEPTION: Warm up time for portable heated rod
. ovens is 30 minutes.
- c. Obtain thermometer or pyrometer from the Calibrated Tool Issue Room (approximate scale range 20*F to 500*F) .
- d. Install thermometer or pyrometer in the oven so that it can be read conveniently. For portable rod ovens, use an adapter block to suspend the instrument in the center o f the oven. Allow 5 minutes exposure for thermometer to stablize before taking a reading.
3 V e. A stationary or portable rod oven is acceptable for use if the temperature is 250*F - 350*F. Unacceptable containers shall not be used until adjusted or repaired and checked in accordance with this procedure.
- f. Make appropriate temperature log entry (Figure 6.9B-1).
3.2.3 Handling Weld Filler Material When "Q" weld filler material is removed from its containers, the material shall be handled and stored to prevent contamination. Adequate protective covering shall be provided for weld filler material remainirg in an opened original container, except as provided for in Section 3.2.1. Contaminated weld filler materials shall be classified as Nonconforming Weld Filler Material (NCWFM) and shall be dispositioned by the PWE in accordance with Section 3.6. When weld filler material is removed from its original container. it shall be identified as follows:
- 1. Each formed consumable insert or individual backing ring shall be marked (imprinted or flag tagged) with the classification, and size; for consumable inserts the heat / lot number is also needed.
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EFFECTIVE O, ' EROWN & R00r, INC. PROCEDURE NUMBER REVISION DATE PACE CPSES JOB 35-1195 CP-CPM 6.9B 2 09/21/84 5 o f 13
- 2. Each 36-inch length of straight bare wire shall be flag tagged on both ends. Marking shall include the material classification. The material shall be stored by classifica-tion, heat / lot, and size.
- 3. Low-hydrogen electrodes shall be segregated to maintain traceability by the classification, size, and heat / lot number when placed in a heated stationary or portable oven.
- 4. Flux-cored and spooled bare wire shall be stored and issued in their containers.
3.2.4 Storage o f Nonconforming Weld Filler Material
- 1. NCWFM shall be stored in a facility or container that is secure by locking device (s) to prevent immediate and casual e ntry.
- 2. NCWD1 containers shall be marked in red and tagged to show the actual or suspected classification and size of the
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material during storage and transit. ._
- 3. NCWFM may be stored in a MDS only for collection and transport to the WQTC for disposition. All NCWFM within a MDS shall be in a marked container (3.2.4, Para. 2) each and all containers shall be maintained in one area within the MDSs and clearly identified as " Nonconforming Weld Filler Material, Do Not Issue".
3.3 ISSUANCE OF WELD FILIER MATERIAL Only "Q" weld filler caterial shall be issued from the MDS and only upon submittal of an approved Weld Filler Material Log (WFML, Figure 6.9B-2) to the MDS attendant. The craf t foreman or weld technician shall enter on the WntL the da te , WP S , filler metal size / class, weld number, welder's symbol, and sign " Issuance Approved" before submittal to the MDS attendant for filler material issuance. In addition, on WFML's to be used on miscellaneous steel the WFML shall contain the drawing identi-fication. EXCEPTION: For Class 2 and 3 support welds which have not been assigned weld numbers, the weld number block shall , be marked "N/A". If for a given weld or welds in such supports a weld number has been assigned prior ,
.. v to welding, the entries shall be made for each l numbered weld. j 1
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'm BROWN & ROM , INC. PROCEDURE NUMBER REVISION DATE PAGE.
CPSES JOB 35-1195 J CP-CPM 6.9B 2 09/21/84 6 o f 13 i NME 1: No more than one weld number or material type per veld ntunber divisions on the WFML is allowed.
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ACCEPTABLE UNACCEPTABLE
.. Weld Date Size / Class Wald Date Size / Class
_ No. No . I 1 1/12/80 3/32 1,2 1/12/80 3/32 1/8 E 70S-2 E 702-2 E 7018 1 1/12/80 1/8 2 1/12/80 3/32 1/8 E 70 t a E70s-2 E 7018 4 j' NME 2: The welder shall be issued a weld technique sheet, j the requested weld filler material, arid the WFML shall
- be returned appropriately completed. ,
I l . NME 3: Any craftsman welding (striking an arc) on a pipe joint l _ shall havs his symbol entered in the space provided on the WFML. h NME 4: When welding is to be done using the automatic CTAW process, the initial entry will be that for the material used for tacking and the welder making the -tack (denoted by "-T" a fter the welding symbol) . A qualified welder j foreman or welding technician may add his symbol to the assigned welders' symbol block and weld with the assigned 4 welders' filler material for a technical demonstration. NME 5: Care should be taken to preclude entries being made l where no welding was accomplished. 1 i NME 6: The use of arrows to denote multiple entries is unaccept-able. f NME 7: Upon issuance, a WDC has a WFML attached having the same serial number as the WDC. This WFML must,be used. Should this WFML become damaged and transcription to a _ blank WFML is required, then the cognizant WT shall ini-tial and date the handwritten transcribed serial number. i NME 8: If no filler material is used on a given weld within a group of welds for which weld filler material has been o issued then this weld number shall be lined through on the WFML by the craftsman, and initialled and dated, for example: If weld material was not used on weld 9 then a line shall be placed through the "9" in the weld number [ ,v'
. column on the WFML.
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BROWN & RCor, INC. PROCEDURE , EFFECIIVE
, REVISION DATE PAGE O '.- NUMBER CPSES JOB 35-1195 CP-CPM 6.9B 2 09/21/84 7 o f 13 NOTE 9: All lined-through entries shall be initialled and dated.
NOTE 10: A quantity of filler rod may be issued for several welds utilizing a MWDC. Upon return to the MDS, the number of rods used may be less than the number of welds. In this case, the total rods for all the welds followed by a "R" may be entered in the. column denoting "Amt RT'd", i.e. : Weld No. Amt. Ant.
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Issued Rt'd 1 2 6R 2 2 6R 3 2 6R 4 2 6R
. Eight rods were issued for welds 1,2,3, and 4. Two rods
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were used to tack these 4 welds. Six rods were returned. m NOIE 11: For Unit 1 and Unit 2, the MDS attendant may issue to l
-V the Craft W"ML's for use on temporary facilities. WFML's
^ for use on Permanent Plant Components will be issued by.
the Package Flow Control . Group. 3.3.1 MDS Attendant Responsibilities The MD Station attendant shall complete the following before issuirg material:
- 1. Verify that Issuance Approval has been given and the WmL has been completed fwWPS, material size and class, welder's
/ symbol, date, and on the NWDC the weld numbers as applicable.
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Enter tWheat number and quantity of material "iss'iTed.' + Q_2. _ s
- 3. Enter the current revision /ICN number to the WPS identified on the WFML. -
NOTE 1: The requirements of this section also- apply to NCWFM .
' in accordance with Section 3.6 with the exception that NCWFM containers shall be marked in red.
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NOTE 2 : For BOP hangers, except those in the mainsteam, stieam clowdown, or feedwater systems, those considered Class 5 or GR, or hangers supporting lines requiring RT, UT, PT or MT the following is acceptable in lieu of (1) above:
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IO BROWN & R00I, INC. PROCEDURE EFFECTIVE CPSES NUMBER REVISICN DATE PAGE JOB 35-1195 , CP-CPM 6.9B 2 09/21/84 8 of 13
- The craft foreman or weld technician shall enter the drawing no., date, size / class, WPS, amount to be issued and. sign " Issuance Approval" prior to presenting to the MDS attendant.
- The MDS attendent will verify the entries made by the craft foreman or veld technician and fill in the ICN No./Rev No. for the defined WPS.
- Upon use in the field the" craft foreman will define Hanger No. (Item No.) in the weld no. column, welders symbol, amount issued to welder, date WPS/
ICN No ./Rev . No . , Heat Io t No . These entries shall be made on the WFML. The " Issuance Approval" blank
- shall be signed by a person on the authorized material issuance list maintained by the MDS. In no case shall the person requesting or using the material be the same person signing for issuance approval.
Weld filler material shall be issued in approved containers
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- .(except constanable inserts and backing rings) and with the follow- ' -
ing amounts as follows:
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- 4. Straight-length bare wire leather pouch identified with a serial ntunber. 40 pieces maximum; 18-inch lengths.
- 5. Spooled bare wire and flux-cored wire on properly marked reels or spools as needed for field applications.
NOIE 3: Weld material received on 25 lbs or 50 lbs. spools may be transferred to approximately 2 lb. spools for issuance. The MDS shall notify Quality Control to witness traceability tra ns fer . ,
- 6. Covered electrodes - Limited to capacity of issuing container.
(Iow-hydrogen electrodes issued in heated ovens only).
- 7. Brazing wire - number leather pouch. 40 pieces maximum; 18-inch length.
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' BROWN & R00r, INC. PROCEDURE EFFECTIVE hafBER BEVISION DATE PAGE CPSES JDB 35-1195 CP-CPM 6.9B 2 09/21/F4 9 of 13 NUIE 4: Consumable inserts and backing rings di not require issuance in an approved container. The material is issued as described on the WFML.
3.3.2 Weld Filler Material - Production Requirements Only one type of filler material may be issued to a welder at a time (C/S, S/S or alloy steel). Two classes of bare or covered rod may not be issued concurrently; for example, ER308 and ER316. Weld filler material shall be used only for the application described on the WFML. Except for immediate use, weld filler material shall remain in
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approved containers. More than one piece of filler material may be removed from the approved container for .short periods when workits conditions require it. the welder shall release the WFML to authorized inspection personnel Corrections to WFML's .nay be made by the personnel d upon request. making the original entry or his supervisor, with such changes being initialed and dated. Fxposure of low-hydrogen electrodes to ambient conditions shall not exceed the followirg time limits: Electrode Classification Maximum Exposure Time (Hours) E 70XX 4 E80XX 2 E90XX 1 E100XX
-- E110XX E3XXXX 4 ak +
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BROWN & RCOI, INC. PROCEDURE EFFECIIVE NUMBER REVISION DATE PACE CPSES v JOB 35-1195 CP-CPM 6.9B 2 09/21/84 10 o f 13 NOIE 1: If filler metal is used only to demonstrate compliance to the applicable WPS in the course of welder surveillance and is not used for production welding the QC Inspector or WT performing surveillance activities shall denote this on the applicable veld filler material log on the same line identifyirg such issuance. NOTE 2: In those cases where an in-process repair is made without documentation on a repair process sheet, and the process used for the repair is different from the last process in the normal welding sequence, it shall be noted on the WFML by a WI that the filler metal was used for an in-process repair. This ensures that an accurate weld history is maintained and that the change is in compliance
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with the WPS. 3.3.3 Weld Filler Material Return m The welder shall return all weld filler material (including stubs),
%) and containers to the MDS/ Supply Room Attendant at the end of the work shift or at the completion of the intended application, whichever occurs first. The WFML shall be maintained with the MWDC or WDC as applicable and returned to the WDS substation at the end of the work shift with the WDC.
NOIE 1: When it becomes necessary to use more than one WFML due to production requirements, the total number of WFML logs used shall be noted on the WDC or MWDC. NOIE 2 : When a constnable insert is issued for a weld and subse-quently damaged and not used, the remnants shall be returned to the MDS to assure that the entry on WFML is lined through for that unused consumable insert. Upon return of weld parameter guide (Figure 6.9D-1), weld filler material and associated container, the MDS/ Supply room attendant
- shall perform the following activities:
- 1. Account for all weld filler material by comparison of the amount received and the amount issued. Make appropriate entries in the "Amo,unt Returned Section."
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s: _ _ _ . ._ __ _ _ BROWN & ROCI, INC. PROCEDURE EFFECIIVE NUMBER REVISICN DATE PAGE
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CPSES
. JOB 35-1195 CP-CPM 6.9B 2 09/21/84 11 o f 13
- 2. Should no filler material be used, the appropriate entry on the WFML should be lined through (cancelled) by the craftsman and initialed and dated. If a weld number is assigned, the weld number should be lined through. If no weld number was assigned for Class 2 and 3 Component Supports, line through the entire entry.
NOTE 3: When a portion of an electrode is returned the total amount shall be noted on the WFML for the total used electrodes. The WFML(s) shall be filed with the applicable WDC when not in use.
- 3.4 IDENTIFICATION OF INCONSISTENCIES When exposed, weld filler material is discovered durirg production welding, the condition shall be described on the WafL: (This is normally done by either the WT, QC Inspector, or CIT.) enter the time of discovery, specify the time of return, initial and date
[9 the entry. Both' the welder and foreman / supervisor shall be informed of the noted condition. Damaged or contaminated weld filler material shall be returned and treated as NCWFM. NOTE 1: This sh 11 also apply to nonfunctioning (broken) portable rod ovens for low-hydrogen electrodes with the exception that:
- a. Welding with materials which have not reached the maximum exposure time may continue.
- b. The weld material and rod oven shall be returned to the MDS station before the exposure limit is reached.
- c. Material returned in "b" above shall be placed in holding ovens and shall not be reissued for at least 8 hours.
Shortages identified by the MDS/ Supply room attendant shall be considered a violation of the weld filler material control program. However, single incidents of minimal shortage (below 5 stubs a day) shall not require corrective action. Repeated shortages by a welder or shortage over 5 stubs a day shall be brought to the attention of the GMS. A shortages log shall be kept .
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J EFFECTIVE O .. BROWN & ROOT, INC. PROCEDURE REVISION DATE PAGE
' ^ CPSES NUMBER w-JOB 35-1195 CP-CPM 6.9B 2 09/21/84 12 o f 13 3.5 DISPOSITION OF RETURNED WELD FILIER MATERIAL (EXCLUDING NCWFM)
Weld filler material returned to the MDS/ Supply room shall be disposi-tioned as follows:
- 1. Consumable inserts and backing rings.
Inserts not used or undamaged may be reissued; all other classify as NCWFM and store for transit to WQTC.
- 2. Straight-length bare wire.
- a. Material not used may be reissued.
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- b. Used material shall be considerd NCWTM.
- c. Flag tag identification must be affixed to all reissued bare wire.
'n Flag tagging shall not be done if material traceability is unclear. - U Material returned without the original flag tag shall be handled as NCWFM.
- 3. Flux-cored wira.
- a. Material not used may be reissued.
- b. Material previously issued may be reissued, if enough exists for further use.
- 4. Iow-hydrogen electrodes.
- a. Material issued in portable oven that was energized (no notation on WFML). Return to stationary oven, reissue as required.
- b. Material issued in portable oven discovered not energized and has exceeded exposure limits (notation on WFML) ,
t disposition as NCWFM, store for transit to WQTC. l l r l
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EFFECTIVE O BROWN & R00r, INC. CPSES PROCEDURE NUMBER REVISION DATE PACE JOB 35-1195 CP-CPM 6.9B 2 09/21/84 13 o f 13
- 5. Other covered electrodes.
Identify and protect, reissue as required.
- 6. Brazing wire.
Under 10 inches, disposition as NCWFM; store for transit to WQTC: 10 inches or over, identify, protect, reissue as required. . 3.6 DISPOSITION / ISSUE /USE OF NCWFM Nonconforming Weld Filler Material shall be kept in controlled storage (either in WQTC, main etorage, or batch plant). Field, use shall be limited to the Batch Plant and Mechanic Shop without specific approval from the PWE. Case-by-case use on Non-Q items may be by written approval of the PWE. Any NCWFM for field use will be issued from the WQTC. 3.6.1 Use of Weld and Brazing Filler Material at the WQTC All weld and brazire filler material stored, issued, and/or used at the WQTC shall be considered NCWFM (except brazing and aluminum filler material) and shall not be removed unless authorized by the PWE or this procedure. All weld and brazing filler material used at WQTC shall be accounted for on the WQTC Filler Metal Use Log (Figure 6.9E-3), but WFML's are not required. However, material issued for use out of the WQTC shall require a WFML. CAUTION: Weld and brazing filler material used at the WQTC for - welder or procedure qualification shall be undamaged, traceable to a GTR and acceptable for use.
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BROWN & R00r, INC. PROCEDURE EFFECIIVE
' NUMBER REVISION DATE PAGE CPSES JOB 35-1195 CP-CPM 6.9B 2 09/21/84 1 of 3 PROCUREMElff 0F WELDING FILLER MATERIAL 1.0 GENERAL Material as supplied, shall be complete in all respects and shall fully conform to the de_scription set forth in the purchase order.
The intent of this section is to secure. material of acceptable weldability and worlananship. All material shall be new and suit-able for the conditions specified. , 1.1 PROCUREMENT FOR SAFETY-RELATED APPLICATIONS
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Materials shall be procured to the requirements established in the applicable' weld filler material specifications referenced in the " Procurement Specifications Index," Table 6.9 BI.
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The GMS shall be responsible for initiating the procurement and maintaining minimum stock levels of weld filler material required V' ~
- 2'" ~r-to meet the Project Schedule..- _
The applicable procurement specification (s), approved by the Project Welding Engineer, shall be attached to Field Requisitions and become part of the purchase order. The initials on the Field Requisition shall indicate PWE review and approval of the re-quisition.
- The requisition and purchase orders shall contain the items list-ed below as a minimum: ;
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- a. Nuclear (Q) Application designation;
- b. ASME-SFA Specificatlon ntsnber and type or classification;
- c. Size;
- d. Quantity;
- e. Testing requirements including chemical analysis, tensil tests,
- Charpy V notch impact properties and testirs temperature. Te n-site and impact tests shall be performed, when required, using specimen from coupons in the as welded and heat treated condi- l t io ns. The time as specified post weld heat temperature shall be 8 hours unless otherwise specified; I
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l I j BROWN & ROCI, INC. PROCEDURE EFFECTIVE O L
^. CPSES NUMBER REVISICN DATE PAGE JOB 35-1195 ' CP-CPM 6.9B 2 09/21/84 2 of 3 PROCUREMENT OF WELDING FILLER MATERIAL (Cont'd)
- f. Test Reports Need Ferrite requirements for austenitic stainless steel filler mate-rials not less than 5%, not more than 12% delta ferrite for wrought steel and not more than 15% for duplex cast steel in undiluted weld deposit;
- f. Packaging Requirements; *
- i. Code addenda to which material must conform
- j. Marking Requirements
- k. Additional special case by case requirements (i.e., specific electrode " Brands" found to be most efficiet.t and of higher quality)
It is mandatory that low-hydrogen-type electrodes be purchased and received in sealed moisture proof containers. Other electrodes and filler materials shall be properly packaged to assure adequate pro-
,.s tection during transit and storage.
lE D N A requirement shall be noted on the field requisition and purchase order that certified materials test reports with actual test results shall be furnished by the welding filler material manufacturer, supplier, or laboratory conducting the tests. Standard certificates of compliance are not acceptable. Identification of all welding materials shall be in accordance with the following ASME Code Section III requirements: " Welding mater-ials shall be clearly identified by legible marking on the package or container to insure positive identification of the material. The marking shall include: heat or lot number a control marking code which identifies the materials with the certified materials test report and other information such as specification, grade, and classification number, supplier's name and trade designation. In accordance with CP-EP-5.0, the Site QA Manager will review and approve field requisitions.
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1.2 PROCUREMENT FOR NONSAFETY RELATED APPLICATIONS , I The GMS may procure material for nonsafety-related applications. '
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The PWE shall review and initial any purchase requisitions.
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,O - BRCWN & R00r, INC. PROCEDURE DATE PAGE CPSES NUMBER REVISION
'-'.s JOB 35-1195 1
CP-CPM 6.9B 2 09/21/84 3 of 3 PROCUREMENT OF WELDING FILIER MATERIAL (Cont'd) Field requisitions prepared for weld filler material intended for nonsafety related applications shall require a Certificate of Compliance to be supplied be fore or upon receipt of the material. The Certificate of Compliance shall attest that the material sup-plied conforms to the requirements of the applicable ASME, SFA, or AWS designation specified in the purchase order. All material procured for nonsafety-related applications shall be treated as nonconforming weld filler material and handled in ac-
- cordance with welding filler material control section of this pro-cedure.
In accordance with CP-EP-5.0, the Site QA Manager will review and p, approve field requisitions. W . j e .1
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EFFECIIVE O BROWN & ROOr, INC. PROCEDURE REVISION DATE PAGE NUMBER
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CPSES JOB 35-1195 CP-CPM 6.9B 2 09/21/84 1 of 2 TABLE 6.9B-1 Weld Filler Material Procurement Specifications Index PROCUREMENT SPECIFICATIONS INDEX NUMBER TITLE WE-010 Mild Steel Covered Arc Welding Electrodes WE-020 Mild Steel Electrodes for Cas Metal Arc and Gas Tungsten Arc Welding WE-030 Iow Alloy Covered Arc Welding Electrodes WE-031 Iow Alloy Covered Arc Welding Electrodes WE-040 Iow Alloy Arc Welding Electrodes for Gas Metal Arc and Gas Tungsten Arc Welding WE-050 Mild Steel Electrodes for Flux-Cored Arc Welding WE-051 Mild Steel Electrodes for Flux-Cored Arc Welding WE-060 Mild Steel Covered (Cellulose) Arc Welding Electrodes WE-070 Corrosion-Resisting Chromita and Chromium-Nickel Steel Welding Covered Welding Electrodes WE-080 Corrosion-Resisting Chromita and Chromium-Nickel Steel Welding Rods and Bare Electrodes - WE-090 Nickel and Nickel-Alloy Covered Welding Electrodes WE-100 Nickel and Nickel-Alloy Bare Welding Rods WE-110 Mild Steel and Iow Alloy Consumable Inserts
^ WE-120 Austenitic Chromium-Nickel Steel Consumable Inserts WE-130 Altninto and Altninum Alloy Welding Rods and Bare Electrodes
- %,,,,/
i
. . . . ~ . .. . . . . .. . . .
r BROWN & ROCI, INC. PROCEDURE EFFECIIVE NUMBER REVISION DATE PAGE CPSES JOB 35-1195 CP-CPM 6.9B 2 09/21/84 2 of 2 Wald Filler Material Procurement Specifications Index (Cont'd) WE-140 Brazing Filler Metal WE-150 Brazing Fluxes WE-160 Flux-Cored Corrosion-Resisting Chromium and Chromium-Nickel Steel Electrodes WE-170 Turgsten Arc Welding Electrodes WE-180 Nickel and Nickel-Alloy Consumable Inserts
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NOTE: QUALITY DOCllMENTS INDEx This list revises and supercedes all previous issues. AM) DATE OF PRINT-0UT _ June 22,1984
,- REV.
In wort 4 DATE DEL. XX Rentsnbered to new procedure - See History Flie
+ Stamped void with no deletion date - See illstory Flie
++ See illstory File HANUAL CODE QAPI B&R Quality Assurance Procedures / Instructions Manual RECV B&R ASME Quality Control Inspection Receiving Manual g
FELO B&R A91E quality Control Inspection Field Manual 1-VOLX B&R ASE Quality Control Inspection Volumetric Examination Manual F01A-85-59 L . 6lly+ 0, 1 i e
?
-n .
e ID DOCUE NT NO. TITLE ORIG. ISS. R EV . DATE IJEL. MANUAL 0000 18 APPENDIX A CPSES QA PROCEDURES ASBREV. LIST QAP
. 19 APPENDIX 8 CPSES QA PROCEDURES GLOSSARY OF TERMS QAP 20 APPENDIX C QA CRITERIA FOR NUCLEAR POWER PLANTS QAP 21 APPENDIX 0 ANSI N45.2 STANDARDS QAP 1 CP-QAP-01.01 QUALITY ASSURANCE PROCEDLRES MANUAL INTRODUCTIUN 07/14/75 0 12/10/81 QAP 158 CP-QAP-01.02 B&R SITE QA ORGANIZATION 07/14/75 +
2 CP-Q AP-02.01 PERSONNEL TRAINING AND QUALIFICATION 06/11/84 11 QAPI 117 CP-QAP-02.01-01 NONDESTRUCTIVE EXAM. AND ECH. INSP. TRAINING, QUAL, & CERT. 01/15/82 XX; QI-QAP-02.UI-01 3 CP-QAP-02.02 STOP WORK 1 01/15/82 XX; CP-QAP-17.02 106 CP-QAP-02.03 QA MANUAL REVISION & CONTROL 06/11/84 4 QAPI 171 CP-QAP-02.04 PROGRAM FOR REPAIR OR ALTERATION OF ASME N-STAMPED COMPONENIS 07/28/83 3 06/15/84 QAPI 4 CP-QAP-03.01 SITE ASME QA ORGANIZATION 06/11/84 6 QAPI
, S CP-QAP-04.01 DESIGN CONTROL 06/11/84 3 QAPI 118 CP-QAP-04.02 SUPPLIER QA PROGRAM REQUIREENTS 07/14/75 08/24/78 115 CP-QAP-05.01 QA REVIEW OF PROCUREMENT DOCUE NTS 07/28/83 4 QAPI 119 CP-QAP-05.02 REVIEW AND APPROVAL OF CONSTRUCTION PROCEDURES 2 12/07/19 XX; CP-QAP-07.01 107 CP-QAP-06.01 PREPARATION OF QA PROCEDIRES & INSTRICTIONS 08/03/83 5 QAPI 120 CP-QAP-06.02 CONTROL OF QUALITT ASSURANCE PROCEDURES MANUALS 0 09/21/78 6 CP-QAP-07.01 CONTROL OF QA PROCED. AND INSTRICTIONS 11/04/83 5 QAPI 121 CP-QAP-07.02 QC RECEIVING INSPECTION 01/04/80 XX; CP-QAP-08.01 122 CP-QAP-07.03 RECEIVING INSPECTION XX 123 CP-QAP-07.04 $UPPLIER DOCUENT REVIEW +
7 CP-QAP-08.01 RECEIVING INSPFCTION 06/11/84 8 QAPI 8 CP-QAP-08.02 EVALUATION AND SELECTION OF SUPPLIES 06/11/84 6 QAPI 9 CP-QAP,08.03 SOURCE SURVEILLANCE
% 06/11/84 5 QAPI 101 CP-QAP-38.04 SUPPLIER AUDITS 06/11/84 4 QAPI Se
]
r, R O ID DOCUE NT NO. TITLE ORIG. 155. REV. DATE DEL. MMUAL CODE 108 CP-QAP-08.05 RECLASSIFICATION OF CODE MATERIAL 06/11/84 4 QAPI 102 CP-QAP-09.01 PEMMENT EQUIPENT TRANSFER VERIFICATION 06/11/84 3 QAPI 124 CP-QAP-10.01 FIELD INSPECTION XX; CP-QAP-il.01
, 73 CP-QAP-10.02 NONDESTRUCTIVE EXAMINATION PROC. QUALIFICATION AND CONTROL 01/27/83 2 QAPI h
109 CP-QAP-10.03 QC SURVEILIANCE OF WELDER PERF0mMCE QUALIFICATION TEST 06/11/84 2 QAPI 10 CP-QAP-11.01 FA8. A INSTALL. INSP. OF C0MPONENTS, 00MPONENT SUPPORTS & PIPING 06/11/84, 5 QAPI, FELD 125 CP-QAP-12.01 CALIBRATION: TOOL CONTROL XX; CP-QAP-13.01 159 CP-QAP-12.01 INSP. CRITERIA & DOCU. REQ. PRIOR TO SYSTEM N-5 CERTIFICATION 06/11/84 11 QAPI, FELD 11 CP-QAP-12.02 SITE SURVEILLANCE N/A 08/25/80 QAP *I
, 160 CP-QAP-12.02 INSP. PROC. & ACCEPTANCE CRITERIA FOR ASE PRESSURE TESTING 06/11/84 8 QAPI FELD 169 CP-QAP-12.03 TESTING FHASE QA FUNCil0NS PRIOR TO ASME CDDE CERT. & STAMPING 08/25/83 3 04/18/84 FELD,QAPI 178 CP-QAP-12.04 IDENTIFICATION, TRACEA8ILITY & RETENTION OF ASE CODE NAMEPLATES 12/28/83 1 06/11/84 FELD, QAPI
, 180 CP-Q AP-12.05 QUALITY ASSURMCE ASME Ill N-5 CERFIFICATION (UNIT 2 ONLY) 01/20/84 0 06/11/84 FELD, QAPI 12 CP-QAP-13.01 CONTROL OF KASURING AND TEST EQUIPE NT 06/11/84 3 QAPI
( l 126 CP-QAP-13.02 CONTROL OF RIGGING ' 01/05/79 XX; 'I
, 105 CP-Q'AP-14.01 INSP. OF STORAGE MD MAINTENANCE OF ECN. EqulPENT 06/11/84 4 QAPI, FELD j 13 CP-QAP-15.01 CONTROL OF INSPECTION STATUS OF ITEMS AND MATERIALS 06/04/82 2 XX; CP-QAP-16.1
. 12/09/82 127 CP-QAP-15.02 IIMDLING OF POTENTIAL REPORTABLE DEFICIENCES 02/27/78 XX; CP-QP-16.01 i
69 CP-QAP-16.01 CONTROL OF NONCONFOMING ITEMS 06/11/84 21 QAPI, VOLX 128 CP-QAP-16.02 REPORTING SAFETV-RELATED DEFECTS & NONC(MPLIANCE 05/21/80 XX; 14 CP-QAP-17.01 CORRECTIVE ACTION 03/01/84 6 QAPI 113 CP-QAP-17.02 STOP WORK lhh ZL QAPI 15 CP-QAP-18.01 PROCESSING OF QA RECORDS 06/11/84 3 QAPI
! 110 CP-QAP-18.02 QA REVIEW OF ASME III DOCUE NTATION 01/27/83 4 06/11/84 QAPI 129 CP-QAP-18.03 SITE AUDITS +
172 CP-QAP-18.03 QA ASE III N-5 CERTIFICATION 06/15/83 1 06/11/84 QAPI
, l t ,h
O q ID DOCUENT NU. TITLE CHIG. ISS. REV. DATE DEL. MANUAL CUDE 16 CP-QAP-18.04 VENDUR QUALIFICATION AND SURVEILLANCE N/A 01/15/82 QAP 98 CP-QAP-18.05 INSPECTION REPORT 06/30/81 0 02/02/83 XX; CP-QAP-II.1 161 CP-QAP-18.06 CONTRUL AND APPLICATION OF ASME CODE SYMBOL SIMPS 06/11/84 3 QAPI 17 CP-QAP-19.00 VENDOR AUDITS 2 01/15/82 QAP. PRCR 111 CP-QAP-19.01 QA AUDITS 12/09/82 1 QAPI 112 CP-QAP-20.01 CONTROL OF CPSES CODE STMPS 01/11/82 0 08/12/82 XX; CP-QAP-18.6 62 QI-HCP-10.01 ECHANICAL EQUIPENT STORAGE MAINTENANCE 1 01/15/82 XX; CP-QAP-14.01 63 Ql-NDEP-101-2 NONDESTRUCTIVE EXAMINATION INSTRUCTION FOR RADIOGRAPHY 3 02/13/81 N/A 64 Ql-NDEP-101-3 CON. CALIB. FILM DENSITY STRIPS & ADJUST. THC DENSIT(NETER 1 02/13/81 N/A 4 65 QI-NDEP-101-5 RADIOGRAPHY FILM ARCHIVAL QUALITT CONTROL 0 02/13/81 N/A 66 QI-NDEP-101-6 INSTRUC. FOR QA SURV. OF RADIO. FILM EVAL. OF ELOS OF SYS. 0 02/13/81 N/A 67 Ql-NDEP-101-7 RADIOGRAPHIC REVIEW FOR MINIMUM WALL PROGRAM 0 02/13/81 N/A
, 68 Ql-NDEP-500-1 ULTRASONIC THICrNESS EASUREMENT 2 02/13/81 N/A 166 QI-NEP-101-4 PERF0%NCE RECORDS +
143 QI-QAP-02.01-01 NONDESTRUCTIVE EXMINAT4UN PExEXNEL CERTIFICATION 03/01/84 6 QAPI N/A 01/15/82 PIM
, 92 Ql-QAP-02.01-03 RECEIVING INSPECTION CERTIFICATION 02/12/81 93 gl-QAP-02.01-04 AUDITOR'S CERTIFICATION 07/14/83 3 QAPI
] 114 Ql-QAP-02.01-05 TRAINING & CERTIFICATION OF ECHMICL INSPECTION PERSONNEL 06/11/84 6 QAPI 22 Ql-QAP-07.01-01 PROCES$1NG OF DOCUENTS BY THE Q4 PROCEDURES COORDINATOR 05/04/82 3 02/02/83 XX; CP-QAP-7.1 165 QI-QAP-07.01-03 ISSUANCE & CONTROL OF QUALITY CONTROL INSPECTION PROC. & INST. 05/04/82 0 02/02/83 XX; CP-QAP-7.1 144 Ql-QAP-07.02-AD RECEIVING OF ELECTRICAL EQUIPENT 1 + 23 QI-QAP-07.02-01 RECEIVING BAGGED CEENT N/A 06/12/80 XX; QI-QAP-08.01-01 1 147 QI-QAP-07.02-02 RECEIVING ELD MATERIAL XX; Q1-QAP-08.01-02 24 QI-QAP-07.02-03 RECEIVING CONERCIAL CROUT N/A 11/12/80 XX; Ql-QAP-08.01-03 148 QI-QAP-07.02-04 RECEIVING E LDING STUDS XX Ql-QAP-08.01-04 l 25 QI-QAP207.02-05 RECEIVING PROTECTIVE C0ATING MATERIAL N/A 11/12/80 XX; QI-QAP-08.01-05 t . 1 -
f O \ \ f ID DOCUK NT NO. TITLE ORIG. ISS. R EV . DATE DEL. HANUAL CODE 26 Ql-QAP-07.02-06 RECEIVING OF BULK MATERIAL-CEMENT AGGREGAIE & ADMIXTURE N/A 11/12/80 XX; Ql-QAP-08.UI-06 149 Ql-QAP-01.02-01 RECEIVING REINFORCEKNT STEEL (RE8AR) XX QI-QAP-08.01-01 150 Ql-Q AP-07.02-08 RECE!VING OF WESTINGHOUSE SAFETY RELATED EQUIPKNT XX QI-QAP-08.01-08 151 QI-QAP-07.02-09 RECE!VING PIPE & FITTINGS XX QI-QAP-08.01-09 152 Ql-QAP-01.02-10 RECEIVING CADWELL SLEEVES & POWDER XX Ql-QAP-08.01-10 a
, i 153 QI-QAP-07.02-Il RECEIVING TVSI/G & H PROCURED SAFETY RELATED IQUIPMENT XX QI-QAP-08.01-Il 154 QI-QAP-07.02-12 RECEIVING STRUCTURAL & MISC. STEEL XX Ql-QAP-08.01-12 27 QI-QAP-07.02-13 RECEIVING RICHMDND SCREW ANCHORS N/A 11/12/80 XX; QI-QAP-08.01-13 !
155 QI-QAP-01.02-14 RECEIVING EASURING & TEST EQUIPENT XX Ql-QAP-08.01-14 28 QI-QAP-07.02-15 RECEIVING MISCELLANEOUS N/A 11/12/80 XX; Ql-QAP-08.01-15 29 Ql-QAP-07.02-16 RECEIVI'iG OF FABRICATED 0881 HANGERS N/A 11/12/80 X X;QI.QAP-08.01-16 i j 145 Ql-QAP-07.02-17 QC RECEIVING INSPECTION XX Ql-QAP-08.01
, 30 Ql-Q AP-08.01-01 RECEIVING BAGGED CEMENT 07/14/83 I QAPI 31 QI-QAP-08.01-02 RECEIVING WELD MATERIAL 08/12/82 5 02/02/83 XX; CP-QAP-8.1 32 Ql-QAP-08.01-03 RECEIVING (DMMERICAL GROUT 03/22/84 2 QAPI 33 Ql-QAP-08.01-04 RECEIVING WELDING STUDS 05/20/82 4 02/02/83 XX; CP-QAP-8.1 r
34 Ql-QAP-08,01-05 RECEIVING PROTECTIVE EDATING MATERIAL 11/12/80 0 XX; CP-QAP-8.1 02/02/83 35 Ql-QAP-08.01-06 RECEIVING OF BULK MATERIAL - CEK MT, AGGREGATE & ADMIXTURES 03/22/84 2 QAPI 36 QI-QAP-08.01-07 RECEIVING RE!!f0RCEMENT STEEL (REBAR) 11/12/80 0 02/02/83 XX; CP-QAP-8.1 37 QI-QAP-08.01-08 RECEIVING OF KSTINGIOUSE SAFETY-RELATED EQUIPKMT 08/12/82 5 02/02/83 XX; CP-QAP-8.1 116 Ql-QAP-08.01-08A RECEIVING WESTINGHOUSE RENEWAL PARTS SUPPLEMENT 08/02/82 1 02/02/83 XX; CP-QAP-8.1 38 QI-QAP-08.01-09 RECEIVING PIPE AMO FITTINGS 08/30/82 7 02/02/83 XX; CP-QAP-8.1 161 Ql-QAP-08.01-09A RECEIVING 8OLTING MATERIAL 08/30/82 1 02/02/83 XX; CP-QAP-8.1 39 Ql-QAP-08.01-10 RECEIVING CADWELD SLEEVES AND POWOER 05/20/82 2 02/02/83 XX; CP-QAP-8.1 40 QI-QAP-08.01-11 RECEIVING TUGC0/G&H PROCURED SAFETY-RELATED EQUIPKMT 08/19/82 5 02/02/83 XX; CP-QAP-8.1 l J 96 QI-QAP'-08.01-12 RECEIVING STRUCTURAL & MISCELLANEOUS STEEL 10/12/82 6 02/02/83 XX; CP-QAP-8.1 O. e e
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i i Q 0 " rN 10 00 CUE NT NO. TITLE ORIG. ISS. REV. DATE IEL. MANUAL C0lE 70 Ql-QAP-08.01-13 RECEIVING RICIN 0ND SCRfW ANCHORS, 01/12/80 0 02/02/83 XX; CP-QAP-8.1 41 QI-QAP-08.01-14 RECEIVING EASURING MD TEST EQUIPENT 09/29/82 3 02/02/83 XX; CP-QAP-8.1 42 QI-Q AP-08.01-15 RECEIVING MISCELLANEUUS MATERIAL 11/12/80 0 02/02/E3 XX; CP-QAP-8.1 43 QI-QAP-08.01-16 RECEIVING OF FA8RICATED CD&I HANGERS 11/12/80 0 01/27/83 XX 44 Ql-QAP-08.01-17 RLVIEW OF DATA REPORTS 08/02/82 4 02/02/83 XX; CP-QAP-S.1 168 Ql-QAP-08.01-18 MATERIAL /EQUIPENT OFFSITE SHIPENTS 08/02/82 0 02/02/83 XX; CP-QAP-8.1 173 QI-QAP-08.01-19 RECEIVING LIQUID PENETRANT MATERIAL 12/09/82 0 02/02/83 XX; CP-QAP-8.1 45 QI-QAP-09.02-048 ULTRASONIC EXAM OF HILTI BOLTS 0 12/18/81 N/A 146 Ql-QAP-10.01-02 INSPECTION OF SITE MISC. & STRUCTURAL STEEL FABRICATION 01/07/80 XX; QI-QAP-11.01-02 163 QI-QAP-10.01-02 WELDER SURVEILLANCE 08/24/78 XX; Ql-QAP-10.03-02 130 Ql-QAP-10.01-03 CONTROL OF SITE FAB. HISC./ STRUCTURAL STEEL MATL. TRACEADILITT 12/28/79 XX; Ql-QAP-11.01-03 K 131 Ql-QAP-10.01-04 WELOING INSP. & FIT-UP OF STAINLESS STEEL LINERS 12/28/79 XX; QI-QAP-11.01-03 132 Ql-QAP-10.01-05_ AMSE SECTION III TRACEABILITY REQUIREMENTS 12/28/79 XX; QI-QAP-11.1-5 l 133 QI-QAP-10.01-20 COMPONENT SUPPORT INSPECTION CRITERIA 01/07/80 XX; QI-QAP-11.1-20 14 Ql-QAP-10.02-01 LIQUID PENETRANT EXAMINATION 02/18/83 3 QAPI,FELO 75 Ql-QAP-10.02-02 MAGNETIC PARTICLE EXAMINATION 07/I5/82 4 QAPI,FELO 76 QI-QAP-10:02-03 RADIOGRAPHIC EXMINATION 02/02/83 4 QAPI, VOLX 77 QI-QAP-10.02-03A NONDESTRUCTIVE EXAMINATION INST. FOR RADIOGRAPHY D0C. 01/27/83 2 QAPI,VOLX 78 QI-QAP-10.02-038 CONTRLG. CAIB. FILM DENS. STRIPS & ADJSTNG. THE DENSIT(METER 01/11/82 1 QAPI,VOLI 19 QI-QAP-10.02-03C RADIOGRAPHIC FILM AROllVAL QUALITY CONTROLS 01/27/83 3 QAPI,VOLX 80 QI-QAP-10.02-030 QA SURVEILEANCE OF RADIOGRAPHIC FILM EVALUATION 02/13/81 1 01/15/82 NUEP 81 QI-QAP-10.02-03E RA010 GRAPHIC REVIEW FOR MINIMlM WALL PROGRAM 02/13/81 0 01/15/82 NDEP 82 Ql-QAP-10.02-04 ULTRA 50NIC THICKNESS EASUREMENT AND LAMINATION EXMINATION 01/11/82 2 QAPI,VOLI , 83 Ql-QAP-10.02-04A ULTRASONIC DIGITAL THICKNESS EASUREENTS 03/01/84 4 QAPI, FELO, VOLX 1 l I y 84 Ql-QAP-10.02-G48 ULTRASONIC EXAMINATION OF HILTI BOLTS 06/02/83 3 01/27/83 J 175 Ql-QAP'-10.02-048 ULTRASONIC EXAMINATION OF HILTI BOLTS 06/02/83 3 QAPI,VOLX f d l il . ..
l r p Q '
~
ID 00 CUE NT R0. TITLE ORIG. ISS. RLV. DATE DEL. MANUAL 00K 179 QI-QAP-10.02-04E ULTRASONIC EXAMIMTION OF ASME SECT, III CLASS 1 PLATE 0 09/20/83 QAPI. VOLX e 85 Ql-QAP-10.02-05 ULTRASONIC EXAMINATION OF ELDENTS 01/27/83 3 QAPI. VOLX i 86 Ql-QAP-10.02-06 LEAK DETECil0N (VACUUM BOX) 05/12/83 2 QAPI 87 Ql-QAP-10.02-07 VISUAL EXAHINATION WELDENTS 02/13/81 6 03/12/82 QAPl. TELD. VOLX 88 QI-YAP-10.02-08 MARKING REQUIREMENTS FOR NONDESTRUCTIVE EXAMINATION 01/11/82 1 QAPI. FELD 99 Ql-QAP-10.02-09 ULTRASONIC EXAMINATION OF ELDS PER AWS Dl.1 01/06/82 1 01/27/83 XX 134 QI-Q AP-10.03-01 QUALITY CONTROL SURVEILLANCE OF ELDER PERFDRMANCE QUAL. TEST + 135 Ql-QAP-10.03-02 WELDER SURVEILLANCE 10/24/79 136 QI-Q4P-10.03-03 WELD MATERIAL SURVEILLANCE . 12/08/79
, 137 QI-QAP-10.03-04 REQUEST FOR REQUALIFICATION OF ELDERS BT ANI 01/04/80 XX; Ql-QAP-ll.1-7 46 QI-QAP-11.01-01 QC SURVEILLANCE OF ELDER PERFURMANCE QUALIFICATION TEST 12/21/19 0 01/15/82 EDI 47 QI-QAP-il.01-02 INSPECTION OF SITE MISC. & STRUCTURAL STEEL FAB. & INSTALL. 0 06/11/81 Mot 48 Ql-QAP-11.01-03 CONTROL OF SITE FAB. MISC./ STRUCTURAL STEEL MATL. TRACEAB. 0 06/11/81 EQt 49 QI-QAP-II.01-05 ASE SECTION III TRACEABILITY REQUIREMENTS 2 01/15/82 EDI 50 QI-QAP-il.01-06 INSPEC. OF THE INSTAL. OF PIPING TO BE EMBEDDED IN Q)NCRETE o 01/08/81 N/A 51 QI-QAP-II.01-07 RFQUEST FOR REQUALIFICATION OF ELDERS BY ANI 01/05/80 0 01/15/82 MOI
- 52 QI-QAP-11.0_l-20 _ INSPECTION OF A9tE COMPONENT SUPPORTS N/A 11/11/79 XX; Ql-QAP-ll.01-28 156 QI-QAP-II.01-21 COLOR CODING AND MARKING OF MATERIALS 01/03/80 157 QI-QAP-l!.01-22 0 EANLINESS CONTROL 01/03/80 53 QI-QAP-11.01-23 QC INSTRUCTIONS FOR PIPE FABRICATION AND INSTALLATION 3 01/15/82 MECH 54 QI-QAP-ll.01-24 INSPECTION OF PRESSURE TESTING 5 01/15/82 ECH 59 QI-QAP-il.01-25 QUALITY ASSURANCE REVIEW 0F ASME III DOCUMENTATION 8 01/15/82 EQt 56 Ql-QAP-ll.01-26 ASME WELD INSPECTION 01/15/82 ++
164 Ql-QAP-11.01-26 ASE PIPE FABRICATION & INSTM.LATION INSPECTIONS 06/11/84 16 QAPI. FELD 57 Ql-QAP ,ll.01-27 INSPECTION OF INSTAL. OF PIPING TO BE EMBEDDED IN CONCRETE O 01/15/82 EOt 58 QI-QAP-II.01-28 FAB. INSTAL. INSPEC. OF ASE CG4PON. SUPPORTS. CLASS 1.2.5 3 06/11/84 25 QAPI. FELD e ., I e
l
~ n o ID DOCUE NF NO. TITLE ORIG. ISS. REV. DATE DEL. MANUAL CODE r
a 170 Ql-QAP-II.01-28A INSTALLATION INSPECTIONS OF A9tE ELASS 1, 2, & 3 SNUUBERS 11/08/83 5 04/18/84 QAPI, FELD 72 Ql-QAP-11.01-29 ACID ETOIING FUSION WELDED PIPE 01/11/82 1 01/27/83 XX 11 Ql-QAP-II.01-30 SPHERICAL BEARING RESTAKING INSPECTION 09/03/82 5 02/02/83 XX; QI-QAP-11.1-20 89 QI-QAP-II.01-31 INST. INSP. OF K CHANICAL JOINTS 12/09/82 5 02/02/83 XX; QI-QAP-11.2-26 , 90 Ql-QAP-II.01-32 INSP. OF SAFETY WIRED FASTENERS 10/01/82 4 02/02/83 XX; QI.QAP-II.1-26, 28 & 28A 94 QI-QAP-11.01-33 litETURNOVER WALKDOWN INSPECTION CRITERIA 06/15/81 0 02/02/83 XX; CP-QAP-12.1 95 QI-QAP-11.01-34 W VALVE LOCATION VERIFICATION 06/15/81 0 01/27/83 XX 97 QI-QAP-11.01-35 PEINAMENT EQUIPENT TRANSFER VERIFICATION 06/26/81 0 01/15/82 XX; CP-QAP-9.1 103 QI-QAP-11.01-36 INSP. VERIFICATION OF SALVAGED SUPPORTS & PARTS 12/09/82 2 02/02/83 XX; Ql-QAP-ll.1-28 104 QI-QAP-11.01-37 INST. QA/QC REVIEW, APPROVAL & UTILIZ. OF CONST. OP. TRAVELERS 02/08/82 1 02/02/83 XX; CP-QAP-ll.1 162 Ql-QAP-II.01-38 FA8, INSTAL!ATION INSP. OF A94E N) MENT RESTRAINTS ELASS I & 2 08/03/83 3 04/18/84 QAPI, FELD 174 QI-QAP-II.01-39 MECHANICAL EQUIPKMT INSTALLATION INSPECTION 06/11/84 4 QAPI, FELD 177 Ql-QAP-II.01-39 A VALVE DISASSEMBLY / REASSEMBLY 06/11/84 3 QAPI, FELD 176 Ql-QAP-II.01-40 INSULATION INST 4.LAT10N INSPECTION 06/11/84 2 QAPI, FELD l 181 QI-QAP-II 02-28 FA8. AND INSTAL. INSP. OF SAFETY CL. COMP. SUPPORTS (U-2 ONLY) 01/20/84 0 04/18/84 QAPI,FELD 138 QI-QAP-12.02-01 SURVEILLANCE OF ASE PIPE & HANGER DOCUMENTATION 08/25/80 139 QI-QAP-13:01-01 QC SURVEILLANCE OF RIGGING H0ISTING 01/05/80 j 140 QI-QAP-15.01-01 FIELD DEFICIENCY - - XX QI-QAP-16.1-1 141 QI-QAP-15.01-02 DOCU. MIN. WALL VIOL. & ARC STRIKE REPAIRS ON NDE REPORT FORM XX QI-QAP-16.1-2 142 Ql-QAP-15.01-04 MATERIAL UPGRADING XX QI-QAP-16.1-4 59 Ql-QAP-16.01-01 FIELD DEFICIENCY REPORTING 3 01/11/82 QAP, MECH, NCR/CA 60 Ql-QAP-16.01-02 DOCUM. BASE KTAL REPAIRS, MIN. WALL VIOL. & ARC STRIKE REPAIRS 05/20/82 4 02/02/83 XX; CP-QAP-16.1 61 QI-QAP-16.01-04 MATERIAL UPGLADING 0 01/15/82 KOI 91 Ql-QAP-16.01-05 DOCUENTING FIELD PIPING DEVIATIONS 03/12/82 3 02/02/83 XX; CP-QAP-16.1 100 QI-QAP-16.01-06 0.05URE OF NCR M-2679 & M-4313 02/02/83 1 09/16/83 QAPI, FELD J y r e
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NUREG-0797 Supplement No. 7 Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446 Texas Utilities Generating Company, et al. U.S. Nuclear Regulatory ; Commission Office of Nuclear Reactor Regulation January 1985 p ra ara vy ag/) s..... 1 9 F01A-85-59 gypy <g Sllli
ABSTRACT Supplement 7 to the Safety Evaluation Report for the Texas Utilities Electric Company application for a license to operate Comanche Peak Steam Electric Sta-tion, Units 1 and 2 (Docket Nos. 50-445, 50-446), located in Somervell County, Texas, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Comanche Peak Technical Review Team of the U. S. Nuclear Regulatory Commission. This Supplement provides the results of the staff's evaluation and resolution of approximately 80 technical concerns and allegations in the areas of Electrical / Instrumentation and Test Programs regarding construction and plant readiness testing practices at the Comanche Peak facility. Issues raised during recent Atomic Safety and Licensing Board hearings will be dealt with in future supplements to the Safety Evaluation Report. 2 Comanche Peak SSER 7 iii
6 TABLE OF CONTENTS P, age A8STRACT ACRONYMS AND ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . vi;
- 1. INTRODUCTION
........ . . . . . . . . . . . . . . . 1-1 The Comanche Peak Special Review Team for SER Supplement 7 . . 1-3 APPENDIX J - Status of Staff Evaluation and Resolution of Technical Concerns and Allegations in the Areas of Electrical / Instrumentation and Test Programs Regarding Constructior Practices at Comanche Peak Steam'klectric Station, Units 1 and 2 . . . . . . J-l
,~
i l - Comanche Peak SSER'7 y
ACRONYMS AND ABBREVIATIONS AA - independent assessment program allegation AB - American Bridge AB - bolt allegation AC - concrete /rebar allegation ACI - American Concrete Institute AD - design of pipe / pipe support allegation ADS - audit discrepancy report AE - electrical allegation AE00 - Office for Analysis and Evaluation of Operational Data (NRC) AFW - auxiliary feedwater system AH - hanger allegation AI - intimidation allegation AISC - American Institute of Steel Construction AM - miscellaneous allegation ANI - authorized nuclear inspector ANS - American Nuclear Society ANSI - American National Standards Institute AO - protective coating allegation AP - pipe and pipe support allegation APC - AMP Product Corporation AQ quality assurance / quality control allegation AQB - QA/QC bolt allegation AQC QA/QC concrete /rebar allegation AQE - QA/QC electrical allegation AQH - QA/QC hanger allegation AQO - QA/QC coating allegation AQP QA/QC pipe and pipe support allegation AQW - QA/QC welding allegation ARMS - Automated Records Management System ASLB - Atomic Safety and Licensing Board ASME - American Society of Mechanical Engineers ASTM - American Society for Testing and Materials AT - acceptance test AT - test program allegation AV - vendor / generic allegation AW - welding allegation B&PVC - Boiler & Pressure Vessel Code B&R - Brown & Root, Inc. BRIR - Brown & Root Inspection Report BRHL - Brown & Root Hanger Locations BRP - Brown & Root piping isometric drawing CAR - Corrective Action Request CASE - Citizens Association for Sound Energy l Ccminche Peak SSER 7 vii ( 4
C&L - Corner and Lada (computer program) C&S - civil and structural CAT - Construction Appraisal Team (NRC) CB&I - Chicago Bridge & Iron Company CCS - Component Cooling 59 stem CEL - Coating Exempt Log CFR - Code of Federal Regulations CHN - construction hold notice CILRT - containment integrated leak rate test CMC - component modification cards COT - construction operation traveler CP - Comanche Peak CP - construction permit CPPE - Comanche Peak Project Engineering CPSES - Comanche Peak Steam Electric Station CPSIG - Comanche Peak Seismic Interaction Group CSTS - Construction and Startup/ Turnover Surveillance Group (TUEC) CVCS - chemical and volume control system CZ Carboline Carbo zine 11 DBA - design basis accident DCA - design change authorization DCC - Document Control Center (TUEC) DCTG - Design Change Tracking Group DCVG - design change verification group DE
- Division of Engineering (NRC)
DFT - dry film thickness DL - Division of Licensing (NRC) D Ameron Dimetcote 6 EDO - Executive Director for Operations (NRC) E&I - Electrical and Instrumentation ETG
- Electrical Test Group (TUEC)
FDSG - Field Damage Study Group (TUEC) FJO - field job orders FP - fire protection FSAR - Final Safety Analysis Report FW - field weld GAP - Government Accountability Project GDC - general design criteria GE - General Electric Corporation GED - General Equivalency Diploma G&H - Gibbs & Hill GHH - Gibbs & Hill hanger (isometric drawing) viii Comanche Peak SSER 7
HFT - hot functional test HIR - hanger inspection report HP - hanger package HVAC - heating, ventilation and air conditioning system HX - heat exchangers IAP - Independent Assessment Program IE - Office of Inspection and Enforcement (NRC) IEEE - Institute of Electrical and Electronics Engineers IM - interoffice memorandum (TUEC) INPO - Institute for Nuclear Power Operations IR - inspection report (NRC) IRN - item removal notice ITT-G - ITT Grinnell JTG - Joint Test Group (TUEC) JUMA - Joint Utility Management Assessment Group LOCA - loss of coolant accident LP - liquid penetrant MAR - maintenance action request M&P - mechanical and piping MCC - motor control center (GE) MDB - master data base MIFI - mechanical fabrication inspector MIL - material identification list (or log) MIME - Mechanical Equipment Inspector MQE - idechanical Quality Engineering MRS - manufacturer's record sheet MWDC - multiple weld data card N/A - not applicable NCR - nonconformance report (TUEC) NDE - nondestructive examination NDT - nondestructive testing NI - never incorporated NONSAT - nonsatisfactory NOV - NPSI - Notice of Violation (NRC) Nuclear Power Service Incorporated NRC - U.S. Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation (NRC) NSSS - nuclear steam supply system O&M - Operations and Maintenance (TUEC) OBE - operating basis earthquake OI - Office of Investigations OJT - on-the-job training Comanche Peak SSER 7 ix
l OL - operating license ORNL - Oak Ridge National Laboratory PC - protective coating PET - permanent equipment transfer PFG - paper flow group PFS - pipe fabrication shop PSAR - Preliminary Safety Analysis Report PSE - Pipe Support Engineering (TVEC) PT - preoperational test PWR - pipe whip restraints P-305 - Carboline Phenoline 305 QE quality engineer QA quality assurance QAI
- quality assurance investigation (TUEC)
QC quality control RCB - Reactor Containment Building RES - Office of Nuclear Regulatory Research (NRC) RFIC - request for information or clarification (B&R) RG - Regulatory Guide (NRC) RI - NRC Region I Office RIR - receipt inspection report (TUEC) RIV - NRC Region IV Office RHRS - residual heat removal system RPI - rod position indication RPS - report process sheet (TUGCO) RPV - reactor pressure vessel RPVI - reactor pressure vessel reflective insulation RRI - Resident Reactor Inspector (NRC) RV - reactor vessel RWN - room work notifications SAP - startup administration procedure SALP - Systematic Assessment of Licensee Performance (NRC) SAT - satisfactory SAVC - structural assembly verification card ; SER
- Safety Evaluation Report (NRC) 1 SIS - Special Inspection Services SMAW - shielded metal arc welding SNM - special nuclear material i SORC - Station Operations Review Committee SRIC - Senior Resiaent Intpector for Construction (NRC)
SRT - Special Review Team (NRC) SSE - safe shutdown earthquake SSER - Safety Evaluation Report Supplement SSPC - Steel Structures Painting Council SSWP - station service water pumps SSI - safe shutdown impoundment Comanche Peak SSER 7 x
STE - system test engineer SWA - startup work authorization TDCR - test deficiency change request TDR - test deficiency report 10 CFR 50 - Title 10 Code of Federal Regulations Part 50 TIDC - TNE - Division TUEC of Technical Nuclear Information and Document Control (NRC) Engineering TP - test program TPD - test procedure deviation Tr - transcript TRT - Technical Review Team (NRC) TSI - thermolag TSMD - Technical Services Mechanical Drafting TSP - tri-sodium phosphate TUEC - Texas Utilities Electric Company , TUGC0 - Texas Utilities Generating Company TUSI - Texas Utilities Service, Inc. UCC - University Computing Company UT - ultrasonic test VC0 - vendor-certified drawing VT - visual weld (inspector) WDC - weld data card WFML - weld filler metal log WPS - welding procedure specification I Comanche Peak SSER 7 xi i _ _ _ _ _ _ _ _ _ _ _ - - - - . - - - - - - -- - -- - ---- - - - - - - - - - - - - ~ - - - - -
1 l 1 ENTRODUCTION On July 14, 1981, the U. S. Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) (NUREG-0797) related to the application by the Texas Utilities Electric Company (TVEC) for a license to operate Comanche Peak Steam Electric Station (CPSES) Units 1 and 2. Subsequently, six supplemental Safety Evaluation Reports (SSERs) were issued by the staff. This report, Supple-ment No. 7, is the first of a series of SSERs dealing with various technical concerns and allegations about construction practices at Comanche Peak. This report addresses approximately 80 technical concerns and allegations in the areas of Electrical and Instrumentation and Test Program. Appendix J to this report provides details of the staff's evaluation and findings of these tech-nical concerns and allegations. The technical concerns and allegations about Comanche Peak were part of the regulatory issues that remained outstanding toward the completion of construc-tion of the Comanche Peak facility. The NRC's Executive Director for Opera-tions (EDO) issued a directive on March 12, 1984, establishing a program for assuring the overall coordination / integration of these issues and their reso-lution prior to the staff's licensing decision. In response to the ED0's directive, a program plan was developed and approved on June 5, 1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of NRC's Region IV Office. This pro-gram plan, entitled Comanche Peak Plan for the Completion of Outstanding Regulatory Actions, specified the critical path issues, addressed the scope of work needed, and provided a projected schedule for completion. Attachment 1 to Appendix J is a listing of the technical concerns and allegations in the aforementioned areas which are grouped according to their areas of discipline. On September 18, 1984, the NRC provided the results of the staff's evaluation of the technical concerns and allegations in the electrical and instrumentation, civil and structural, and test program areas, identifying potential safety con-cerns and requesting additional information, including a program and schedule for completing a detailed and thorough assessment of the concerns identified. (See Attachment 3.) This requested information was submitted by TUEC on October 8, 1984, in the form of a proposed program plan. TUEC has partially revised this program plan in a letter to NRC of November 21, 1984. The revised program plan, once approved, as well as its implementation, will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak Unit 1. Attachment 2 to Appendix J provides the staff's detailed assessment of the individual technical concerns and allegations in the electrical and instrumentation and test program areas. Management and coordination of all the outstanding regulatory actions for Comanche Peak are under the overall direction of Mr. Vincent S. Noonan, the NRC Comanche Peak Project Director. Mr. Noonan may be contacted by calling 301-492-7903 or by writing to the following address: Comanche Peak SSER 7 1-1 n
Mr. Vincent S. Noonan Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Copies of this Supplement are available for public inspection at the NRC's 20555, and the Public Document Room at 1717 H Street, NW, Washington, D. C. Local Public Document Room, located at the Somervell County Public Library On Availability of all The Square, P. O. Box 1417, Glen Rose, Texas, 76043. material cited is described on the inside front cover of this report. Comanche Peak SSER 7 1-2 I l
l The Comanche Peak Technical Review Team for SER Supplement 7 Bsahm, D. - Brown, C. Idaho National Engineering Laboratory EG&G, San Ramon Calvo, J. - Gagliardo, J. - Office of Nuclear Reactor Regulation, NRC j Haughney, C. Reactor Training Center, IE, NRC COMEX Corporation Hofmayer, C. - Ippolito, T. - Brookhaven National- Laboratory Office of Analysis and Evaluation of Operational Data, NRC
- Johnson, A. -
Region IV NRC
- Kaimig, R. -
Region I, NRC Laudenbach, D. - EG&G, San Ramon Li, H. - Mackley, A. - Office of Nuclear Reactor Regulation, NRC Marini, W. - Idaho National Engineering Laboratory Marinos, E. - Resource Technical Services, Inc. Myers, G. Office of Nuclear Reactor Regulation, NRC Parameter Noonan, V. - 011u, W. - Office of Nuclear Reactor Regulation, NRC Division of Technical Information and Document Control, NRC Poslusny, C. - Office of Nuclear Reactor Regulation, NRC i, Selan, J. - Smith, W. Lawrence Livermore National Laboratory Region IV, NRC Tang, R. C. - Vietti, A. - Office of Nuclear Reactor Regulation, NRC Wassman, R. - Office of Nuclear Reactor Regulation, NRC White, R. - Office of Nuclear Reactor Regulation, NRC 1 Lawrence Livermore National Laboratory i 4 1 l 4 i 1 Comanche Peak SSER 7 1-3 \ 1 7
i i d i I i j APPENDIX J STATUS OF STAFF EVALUATION t AND RESOLUTION OF TECHNICAL CONCERNS AND ALLEGATIONS IN THE AREAS OF ELECTRICAL / INSTRUMENTATION AND TEST PROGRAM REGARDING CONSTRUCTION AT COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 4 .l i I
~,
t 4 i i .
TABLE OF CONTENTS P_ age
; 1. Introduction ......................... J-1
- 2. Comanche Peak Technical Concerns and Allegations Management Program . . . ........................
J-3 2.1 Background . ....................... J-3
- 2. 2 Review Approach and Methodology ............. J-3 2.2.1 Concern and Allegation Tracking System ...... J-3 2.2.2 Review Methodology ................ J-4 2.2.3 Interviews with Allegers ............. J-5 2.3 Communications with TUEC . . . . . . . . . . . . . . . . . J-5
- 3. Summary of Evaluations
.................... J-7 3.1 Electrical and Instrumentation Group Summary . . . . . . . J-7 4
3.1.1 Scope of Concerns and Allegations . ........ J-7 3.1.2 Electrical and Instrumentation Group ....... J-8 3.1.3 Findings for Electrical and Instrumentation Issues ...................... J-8 1 3.1.4 Overall Assessment and Conclusions ........ J-10 3.2 Test Program Group Summary . . . . . . . . . . . . . . . . J-11 3.2.1 Scope of Concerns and Allegations . . . . . . . . . J-11 3.2.2 Test Program Group ................ J-13 3.2.3 Findings for Test Program Issues ......... J-13 3.2.4 Overall Assessment and Conclusions ........ J-14 4 Actions Required of TUEC ................... J-15 4.1 Electrical and Instrumentation Area. . . . . . . . . . . . J-15 4.1.1 Electrical Cable Terminations . . . . . . . . . . . 4.1.2 J-15 Electrical Cable Tray and Conduit Installations . . J-15 4.1.3 Electrical Equipment Separation . . . . . . . . . . J-16 4.1.4 Control Room Ceiling Fixture Supports . . . . . . . J-16 4.1.5 Electrical QC Inspector Training / Qualifications . . J-16 l Comanche Peak SSER 7 J-iii l __
._ _ . - _ ~ - - - - - ~ - - - - - - - - - - - - - - - - - - -
TABLE OF CONTENTS (continued) Pa!Le J-17 4.2 Test Program Area .................... 4.2.1 Hot Functional Testing .............. J-17 J-17 4.2.2 Containment Integrated Leak Rate Testing J-17 4.2.3 Prerequisite Testing ............... J-18 4.2.4 Preoperational Testing .............. Attachments Attachment 1 - Listing of Technical Concerns and Allegations in the Electrical and Instrumentation and Test Program Areas. Attachment 2 - Assessment of Individual Technical Concerns and Allegations in Electrical and Instrumentation and Test Program Areas. Attachment 3 - September 18, 1984, letter with enclosure, D. G. Eisenhut, i Director, Division of Licensing, Office of Nuclear Reactor Regulation, NRC, to M. D. Spence, President, Texas Utilities Electric Company, subject: Comanche Peak Review. Attachment 4 - October 5, 1984, letter with enclosure, D. G. Eisenhut, Director, i Division of Licensing, Office of Nuclear Reactor Regulation, NRC, to M. D. Spence, President, Texas Utilities Electric Company, subject: errata sheet for September 18,.1984, letter. J-iv Comanche Peak SSER 7
I i i i i i
- 1. Introduction ;
As construction of the Comanche Peak Steam Electric Station was nearing ] j completion, issues that remained to be resolved prior to the consideration of issuance of an operating license were complex, resource intensive, and spanned more than one NRC office. To ensure the overall coordination and integration j i of these issues, and to ensure their resolution prior to licensing decisions, the NRC's Executive Director for Operations (EDO) issued a memorandum on March , i
; 12, 1984, directing the NRC's Office of Nuclear Reactor Regulation to manage ; all necessary NRC actions leading to prompt licensing decisions, and assigning the Director, NRC's Division of Licensing, the lead responsibility for coordina-ting and integrating the related efforts of various offices within the NRC.
i The principal areas needing resolution before a licensing decision on Comanche Peak can be reached include: (1) the completion and documentation of the staff's review of the Final Safety Analysis Report (FSAR); (2) those issues in contention before the NRC's Atomic Safety and Licensing Board (ASLB); (3) the i completion of necessary NRC regional inspection actions; and (4) the completion t and documentation of the staff's review of technical concerns and allegations a regarding design and construction of the plant. j
! Technical concerns and allegations about Comanche Peak, totalling approxi-mately 600, have been raised mainly by the quality assurance / quality control (QA/QC) personnel working or having worked on site. Their job responsibili-j ties involve or involved QA/QC aspects of safety-related structures, systems, ! and components to determine whether and to what extent such items are manufac-tured, purchased, stored, maintained, installed, tested, and inspected as re-quired by project documents and procedures. Many of these allegations were ; made orally to NRC Region IV staff, NRC Comanche Peak Site Resident Inspectors, ! NRC investigators, or in letters to the NRC, as well as in testimony before the Atomic Safety and Licensing Board (ASL8). Individuals with allegations were 1 also sponsored by the intervenor group Citizens Association for Sound Energy (CASE) and the Government Accountability Project (GAP). General allegations about poor construction work at Comanche Peak were also made in several news- ] paper articles in the Gallas/ Fort Worth, Texas areas.
By the end of April 1984, the staff identified approximately 400 technical concerns and allegations related to the construction of the Comanche Peak facility, including findings by NRC's Special Review Team. (See Section 2.1 !' below.) During its investigation of a concern or allegation, the TRT identi-fled additional concerns. Interviews with allegers also yielded additional concerns. By December 1984, approximately 600 concerns and allegations had been identified. These technical concerns and allegations were grouped by subject into the follow- ! ing areas: 1 1 - Electrical and Instrumentation i - Civil and Structural Mechanical.and Piping 1 1 j Comanche Peak SSER 7 J-1 i
l l l i
- Quality Assurance and Quality Control (QA/QC)
- Coatings
- Test Program Miscellaneous This report is the first of a series of reports dealing exclusively with the NRC staff's efforts to evaluate and resolve the technical concerns and allega-tions raised by various parties and individuals regarding construction practices at the Comanche Peak facility. An allegation or concern was assessed as having no safety significance if, based on technical findings, the assessment showed that a structure, component, or system would perform its intended function.
Subject areas covered in this report include electrical and instrumentation and test program. The technical concerns and allegations in the areas of civil and structural, mechanical and piping, coatings, QA/QC, and miscellaneous issues, as well as the remaining areas of outstanding regulatory actions, will be ad-dressed in future supplements to the Comanche Peak Safety Evaluation Report (SER). The staff's findings for electrical and instrumentation and test programDetails allegations or concerns are summarized in Section 3 of this Appendix. of the assessment and findings on individual concerns or allegations appear in Attachment 2 to this Appendix. Those aspects of the concerns or allegations that pertain to wrongdoing (e.g. , falsification of records) were forwarded to the NRC's Office of Investigations (OI) for followup bechuse they are outside the scope of the technical staff's review. A number of potential violations of NRC rules and regulations have been identi-fied during the course of the TRT investigation. These potential violations have not been addressed in this SSER, but will be further reviewed by the NRC Region IV staff, which will determine appropriate followup actions. Comanche Peak SSER 7 J-2
_ _ . _ . . _ . _ - . _ _ _ _ _ ~ . _ _ _ _ . . _. _ _ . _ _ . ._ _ _. _ . _ . _ 1 l
- 2. Comanche Peak Technical Concerns and Allegations Management Program
- 2.1 Background I Shortly after the E00's issuance of the March 12, 1984, directive, the staff 1
found it necessary to (1) obtain current information relative to TUEC's i management control of the construction, inspection, and test program and (2) ! obtain necessary information to establish a management plan for resolution of i i all outstanding licensing actions. In order to achieve these goals in an expeditious and objective manner, a Special Review Team (SRT) was formed to conduct an unannounced review of the Comanche Peak plant. The SRT consisted j of eight reviewers and one team leader, all from NRC's Region II Office, and a team manager from NRC headquarters. The SRT spent over 800 hours, from April 3 to April 13, 1984, performing this review. The SRT concluded that TUEC's programs were being sufficiently controlled to allow continued plant construc-tion while the NRC completed its review and inspection of the Comanche Peak i facility. The SRT review also provided a basis for the development of an NRC management j plan for the resolution of all outstanding licensing actions. This plan was
) approved on June 5, 1984, by the Directors of NRC's Office of Inspection and i
Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of i NRC's Region IV Office. The purpose of the plan was to ensure the overall i coordination and integration of the outstanding regulatory actions at Comanche Peak and their satisfactory resolution prior to a licensing decision by the
- NRC. In accordance with the plan, a Technical Review Team (TRT) was formed to evaluate and resolve technical issues and those allegations that had been identified. On July 9, 1984, the TRT begt.n its 10-week (five 2-week sessions) j onsite effort, including interviews of allegers and TUEC personnel, to determine 4 the validity of the technical concerns and allegations, to evaluate their safety significance, and to assess their generic implications. The TRT consisted of about 50 technical specialists from NRC headquarters, NRC Regional Offices, and NRC consultants, who were divided into groups according to technical discipline.
! Each group was also assigned a group leader. l 2.2 Review Approach and Methodology i
- 2.2.1 Concern and Allegation Tracking System i
4 A tracking system was developed for identifying and listing each concern or allegation. These technical concerns and allegations were grouped according to their topical areas or disciplines, and were listed numerically within each
- group in the order that they were identified by the TRT. The tracking system
- included a description of the concern or allegation; its status or the actions j
taken to resolve it; the nature of the sources of the concern or allegation j (i.e., anonymous or confidential); a code for the individual who identified the concern or allegation (instead of the individual's name); the date when the concern or allegation was received by the TRT; the source document (e.g. , j letter, NRC inspection report, hearing transcript, etc.); cross reference; etc.
; At the end of each 2-week session, the concern / allegation tracking system was
! updated, as needed, to reflect the status of each concern or allegation, as j well as any new ones that had been added. Comanche Peak SSER 7 J-3 i
i f i 4 1 2.2.2 Review Methodology The technical concerns or allegations similar in subject were combined and i evaluated as one category. For each concern / allegation or concern / allegation j-category, an approach to resolution was developed by the cognizant reviewer (s).
~
Each approach to resolution was reviewed cnd approved by the responsible group leader. The group leaders and reviewers were instructed to: i
- develop and maintain a work package for each issue or category of issues that contained or referenced pertinent documentation associated with the
{ issue (s) and the ultimate resolution, including records of interviews and i ; j inspections for supporting the final NRC staff decisions regarding the ; issue (s); and to ^ protect the identity of the allegers, as a matter of NRC practice. Such l efforts included limited and controlled distribution of allegation-related j documentation and correspondence; minimal use of names, identifying titles, or position descriptions in written material; enlarged sampling l i of activities to prevent direct links by non-NRC personnel between the activity under investigation and the alleger; and other indirect approaches toward investigating the allegations. During TRT onsite sessions, daily meetings were held at the review group level
< to assess progress, to adjust the inspection and evaluation approach as needed, and to provide a forum for the reviewers to interact withSimilar one another daily or to meetings discuss-problems and to arrive jointly at resolutions.
l were also held at the management level where the group leaders interacted with one another and with the Project Director, his assistant and staff. j In evaluating the technical concerns and allegations, the TRT reviewers i examined areas in the plant where direct observation could provide information 4 needed for evaluating an allegation or concern. -During its onsite sessions, j the TRT interviewed the allegers as needed to clarify their concerns or allega- '
- tions. To the extent possible, the TRT contacted allegers after its onsite -
review to discuss preliminary TRT findings and to obtain any additional comments ' from them. (See Section 2.2.3 below.) The TRT also. interviewed TUEC and TUEC contractor personnel as was warranted by the evaluation. In addition to these
- contacts, the TRT reviewed various project documents, including specifications, i
engineering drawings and analyses, procedures, instructions, NRC Region IV i inspection reports, and applicable sections of the Final Safety Analysis
' Report (FSAR) and NRC regulations pertinent to the allegation or sample selected by the TRT for inspection. The TRT also examined construction records, such as design change authorizations, construction work packages, QC inspection reports, nonconformance reports, deficiency logs,. lists and reports, and QC inspector training and certification records. In addition, the TRT i
j l reviewed pertinent transcripts from recent ASLB hearings and depositions of
; TUEC personnel and former employees.
Based on these reviews and interviews, the TRT determined the validity of each technical concern or allegation and assessed its safety significance, its potential generic implications, and any indications of potential management breakdown. Detailed documentation of the TRT assessment and final determina-
' tions of each technical concern or allegation appear in Attachment 2 to this Appendix.
Comanche Peak SSER 7 J-4 4 : i :
2.2.3 Interviews with Allegers Approximately 600 technical concerns and allegations regarding the construction of the Comanche Peak facility have been raised by approximately 70 allegers through various mechanisms. During its onsite work, the TRT interviewed 18 individuals in person, some of whom received followup interviews by telephone. For ten allegers, the TRT reviewers were able to obtain the needed information by telephone and determined that personal interviews would not be necessary. Three allegers contacted by the TRT declined being interviewed. Five allegers couid not be located during the TRT's onsite sessions because their current addresses and telephone numbers were not available. They have not responded to correspondence from the TRT sent to their last known addresses expressing the TRT's intention to discuss their concerns with them. Efforts to locate these individuals included inquiries through the NRC's Office of Investigations, NRC's Region IV staff, the telephone company and U.S. Postal Service, selected inquiries of their relatives and former co-workers, confidential examination of the personnel files of TUEC and its contractors, and in some cases, inquiries to the intervenor group, the Citizens Association for Sound Energy (CASE), and the Government Accountability Project (GAP). To the extent possible, the TRT kept a transcript for each personal interview conducted during its onsite sessions. The names and identities of the allegers had been deleted from the transcripts, as well as from other pertinent reference or source documents, before TRT reviewers were given any portions of these documents for review and follow-up. During the TRT's onsite work, the original transcripts were kept in a locked file in the TRT Project Director's office. The distribution of these transcripts within the NRC, and even within the TRT, was limited and controlled. Subsequent to its onsite work, and at the completion of its evaluation, the TRT attempted to contact each alleger to discuss the TRT's findings regarding their original concerns, and to obtain additional comments from them, if any. Thirty allegers have received such followup interviews. A total of 19 allegers could not be located. Some of these individuals had received initial TRT interviews but had since left the area. Three allegers declined to have further contacts with the TRT. Interviews with the remaining allegers are planned during January and February of 1985. The outcome of followup interviews conducted through December 1984, is briefly discussed in the individual SSER sections in Attach-ment 2. Transcripts were kept for all followup interviews conducted either by telephone or in person. 2.3 Communications with TUEC Whenever the TRT reviewers encountered problems during their evaluations, the TRT Project Director and/or his designee resolved them through discussions with TUEC management onsite. There were also frequent staff-level contacts between TRT members and TUEC personnel during the TRT's onsite activities. In keeping with the NRC practice of promptly notifying applicants of outstanding information/ evaluation needs that could potentially affect plant safety, the staff held several meetings with TUEC representatives at NRC headquarters toward the end of the TRT's review. These meetings were held to discuss potential safety concerns and to request additional information needed by the TRT to complete its review, Comanche Peak SSER 7 J-5
The NRC staff met with TUEC representatives for the first of these meetings on September 18, 1984, to discuss TRT findings for electrical and instrumentation, civil and structural, and test program allegations and concerns. A letter documenting these findings and a request for additional information was issued to TUEC on the day of the meeting (Attachment 3). TUEC later submitted the requested information in the form of a proposed program plan, delineating planned actions to address the deficiencies identified by the TRT. The TRT met with TUEC representatives to discuss this proposed program plan on October 19 and 23, 1984. TUEC submitted a partially revised program plan to NRC on November 21, 1984. On November 29, 1984, NRC sent a letter to TUEC containing potential open issues and requesting additional information and proposed program plans for mechanical and piping and miscellaneous allegations and concerns. The letter also provided TUEC with the status of NRC's evaluation of coatings allegations. Informal telephone discussions between TRT group leaders and their TUEC counterparts regarding these letters have been ongoing. (Reports documenting these discussions have been made available to CASE and are available for inspection at the NRC Public Document Room, 1717 H St., N.W., Washington, D.C. 20555, and at the Comanche Peak Local Public Document Room, Somervell County Public Library On The Square, P.O. Box 1417, Glen Rose, Texas 76043.) On January 8, 1985, the NRC issued a letter to TUEC informing them of the TRT's findings in the construction QA/QC area and requesting a program and schedule for completing a detailed and thorough assessment of the QA issues presented in the letter. A meeting between TUEC and the TRT wasTUEC's held on January 17, 1985, to discuss potential open issues in the QA/QC area. proposed program plan for each of the subject areas and its implementation of the plan will be evaluated by the NRC staff prior to the NRC licensing decision on Comanche Peak. l l Comanche Peak SSER 7 J-6 r l
l l
- 3. Summary of Evaluations 3.1 Electrical and Instrumentation (E&I) Group Summary 3.1.1 Scope of Concerns and Allegations The concerns in the E&I area relate to construction activity including equipment installation, specifications, drawings, procedures, personnel training and quali-fication records, and inspections. There are 53 concerns and allegations in this area, 20 of which are hardware related and 33 of which are QA/QC related. The E&I Group reviewed an additional item of concern identified by the SRT regarding overloading of cable trays due to the installation of "thermolag" material. The above concerns and allegations were consolidated by subject into nine separate categories. A concern or allegation may have been assigned to several appli-cable categories if it raised issues that were common to the subject categories.
When assigning QA/QC-related allegations to subject categories, those with available information on specific equipment location were also assigned to the hardware-related categories such that a direct inspection of the equipment installation involved would be performed. The nine categories and their characterizations are as follows: Category No. Subject Characterization of Concerns and Allegations 1 Electrical Cable Improper-sized lugs, improper use of cable Terminations butt splices in panels and cable termina-tions not conforming with drawings. 2 Electrical Cable Problems with cable tray seismic supports, Tray & Conduit clearance of process pipes from cables in Installation cable trays and loose conduit fittings. 3 Electrical Equipment Violation of the cable separation criteria Separation between separate cables, trays and conduits and inconsistency between specifications and regulatory requirements. 4 Control Room Ceiling Field run conduit, drywall and lighting
- Fixture Supports supports in the control room classified as non-seismic.
5 Electrical Improper generation and disposition of Noncomformance electrical NCRs. Report (NCR) Activities 6 Electrical QC Inspectors inadequately qualified and Inspector received help to pass certification Training and tests. Qualifications 7 Electrical Cable Cable tray overfill, cable spliced Installation in trays and improper cable dressing. Comanche Peak SSER 7 J-7
8 Electrical Precedures Changes requirements for electrical inspection procedures without proper justification. 9 Electrical Inspection Inspection reports written without Reports, Inspection reinspections; in process inspections Item Removal not conducted. Notices and In-Process Inspections 3.1.2 Electrical and Instrumentation Group The E&I Group consisted of seven reviewers who, collectively, represent 140 years of engineering experience, of which 90 years were in the nuclear industry in electrical and instrumentation engineering design, quality assurance and control, inspection, construction, project management and regulatory activi-ties. The E&I Group members included two representatives from the Office of Nuclear Reactor Regulation, one from the NRC Region IV Office, three from a national laboratory, and two from consulting firms. 3.1.3 Findings for Electrical and Instrumentation Issues i Each of the E&I SSER categories lists and characterizes all the concerns raised in the allegations and by the Special Review Team. In some instances, the E&I Group, during its evaluation of an allegation, discovered a new concern unre-lated to the original allegation. These new concerns were also evaluated and reported in the appropriate category. An assessment of the safety significance of the concerns, as well as the generic implications of the findings and the root cause of each situation, as appropr-iate, are also presented. In addition category includes conclusions, staff positions, and actions required of TUEC. On September 18, 1984, the TRT presented at a public meeting the E&I findings, as well as the actions required by TUEC to reach final resolution of the issues. The TRT noted at that meeting that the E&I findings, as well as the actions required of TUEC, could not be considered final until they were integrated with the results of the overall programmatic review being conducted by the QA/QC Group. Since then, minor modifications were made to these findings and actions to include the results of the review of additional information and to integrate them with the results of the review by the QA/QC Group. The QA/QC areas involved are referenced in the E&I categories. The E&I Group found no problems with the concerns raised by the allegations or the Special Review Team regarding the installation of electrical cables; nor could the E&I Group find any evidence of discrepancies in the electrical NCR activities, electrical procedures, electrical inspection reports, inspection removal notices and in process inspections. The E&I Group concludes that the ' concerns in these areas either could not be substantiated or have no safety l significance with respect to the items identified. Comanche Peak SSER 7 J-8 l
In the cable terminations area, the E&I Group found problems with the installation and inspection procedures and documentation of butt splices in panels; the documentation of safety-related and associated terminations in panels; and the disposition of NCRs related to vendor-installed terminal lugs. The E&I Group concludes that there are concerns about the adequacy of TUEC's QC inspection program. (See Attachment 2, E&I Category 1.) The E&I Group found only one problem in the installation of electrical cable tray and conduit: craft personnel lacked training in the use of an installa-tion manual for conduit and junction box supports. (See Attachment 2, E&I Category 2.) In the area of electrical equipment separation, the E&I Group found several cases of separate safety- and nonsafety-related cables and flexible conduits (containing safety- and nonsafety-related cables) inside main control room panels that did not meet minimum separation requirements. The TRT found no evidence to justify this lack of separation. The E&I Group found two instan-ces of violation of the separation criteria concerning separation of redundant instrummtation and field wiring by barrier. The E&I Group also found that TUEC's existing analysis substantiating the acceptability of the criteria for separation between independent conduits and cable trays had not been reviewed by the NRC staff. The E&I Group therefore concludes that there are concerns about the adequacy of TUEC's QC inspection program. (See Attachment 2, E&I Category 3.) The potential safety significance and generic implications concerning the control room ceiling fixture supports was jointly reviewed by the E&I and the civil and mechanical Groups. Regarding the electrical aspects of this concern, the E&I Group concludes that the installation of the nonsafety-related conduit in the control room was inconsistent with seismic requirements and that the suspended drywall ceiling and lighting supports appeared to satisfy seismic requirements, but no analysis could be found that confirmed the adequacy of the supports. The E&I Group also inspected selected seismic Category I areas of the plant and concludes that the installation of nonsafety related conduits of less than or equal to 2 inches in diameter is inconsistent with seismic installation requirements. (See Attachment 2, E&I Category 4.) The last issue of potential safety significance concerned the lack of programmatic control of the electrical QC inspector qualification program, which may be indic-ative of inadequate qualification for some electrical QC inspectors. Since the training and certification program is the same for all disciplines (except ASME), the E&I Group concludes that the deficiencies identified with the electrical QC inspector training and qualifications may have implications for other con-struction disciplines. The implications of the E&I Group findings were further assessed by the TRT QA/QC Group as part of the overall programatic review of QC inspector training and qualification. (See Attachment 2, E&I Category 6; also see QA/QC Category 4, " Training and Qualification.") i Comanche Peak SSER 7 J-9
The E&I findings and actions required by TUEC (presented in Section 4 of this SSER) as related to the specific concerns and allegations were discussed with those individuals responsible for raising them and willing to participate in these discussions. Any disagreements with the E&I findings noted by these individuals, as well as the E&I Group resolutions concerning them, are reported in the appropriate E&I category. 3.1.4 Overall Assessment and Conclusions Most of the concerns and allegations were raised by electrical quality control (QC) inspectors and were found to be very general, and often without any specific connection between the concern and plant safety. These problems were apparent in several of the concerns and allegations addressing problems with nonsafety-related equipment. Further contact with the individuals raising the concerns did not provide the required specificity to focus on the concerns. The general nature of the concerns and the absence of specific exploration of safety signif-icance of the concern may be an indication of lack of proper training in elec-trical QC inspection, even though some QC inspectors had experience on this type of work at nuclear power plant facilities other than CPSES. In general, the quality of the E&I installations reviewed by the E&I Group was found to be acceptable, except for those cases which the E&I Group determined to have safety significance. To determine the extent of the generic implication of these concerns, TUEC is required to conduct further review and inspections. (See Section 4, below.) The E&I Group concludes that the problems found with electrical cable termina-tions, electrical equipment separation and control room ceiling fixture sup-ports, together with the findings concerning inadequate training and qualifica-tion of electrical QC inspections, are an indication of programmatic weakness in QC. The deficiencies identified during the E&I review of both hardware installation and QA/QC-related matters indicate weaknesses in the QA/QC program and are con-sidered in the overall programmatic review by the QA/QC Group. The QA/QC pro-grammatic review will consider the breadth and depth of the actions required by TUEC to resolve not only the specific E&I concerns identified in this report, but also other programmatic concerns related to construction activities of E&I installations. Therefore, the E&I Group concludes that any actions taken by TUEC to resolve the specific E&I concerns identified, or to establish root causes and appropriate corrective actions concerning them, should not be con-sidered final until they are properly integrated with the results of the pro-grammatic review performed by the QA/QC Group. Comanche Peak SSER 7 J-10
l l 3.2 Test Program (TP) Group Summary I l 3.2.1 Scope of Concerns and Allegations i The technical concerns and allegations in the Test Program area involve the i prerequisite and preoperational testing phases for CPSES Unit 1. There were i a total of 18 determined byconcerns the TRT. and allegations in the Test Program area, as originally (Several closely related allegations were combined.) j Thirteen of these were contained in a proposed contention (No. 26) proffered by the Citizens Association for Sound Energy (CASE) on October 13, 1983, to 1 the Atomic Safety and Licensing Board (ASLB) sitting in the Comanche Peak operating license hearing. While the proposed contention was ultimately j not admitted by the ASLB, the technical concerns expressed by CASE were con-sidered by the TRT in its evaluations. The remaining five allegations were brought forward by the Government Accountability Project (GAP) and CASE, which 3 had received them from a confidential source during conversations, and later in i 9.he form of an affidavit. The TRT reviewed the affidavit and pursued i information from it. l The technical concerns and allegations in the Test Program area were catego-1 rized into the following seven general topics: i Category t Number Subject Characterization of i 1 Concerns and Allegations ' 1 Hot Functional Testing HFT was deficient in that not all (HFT) components and equipment were { installed at the time of testing; neither TUEC nor the NRC Region IV staff noticed this condition; i neither kept the ASLB informed of l the problems encountered during i HFT; TUEC and the NRC Region IV i staff were willing to accept deficient test results; the HFT did not take accident conditions , into consideration; TUEC and the NRC Region IV staff were willing ' i j to accept deficient test results, , 2 Unit 2 Testing Although the NRC requires that each unit at a multi-unit site undergo a test program which complies with Regulatory Guide l 1.68, TUEC would not conduct a test program on Unit 2, but rather would rely on the results ; of Unit 1 testing, unless other- , wise ordered by the ASL8. i t Comanche Peak SSER 7 J-11 t
Category Characterization of Subject Concerns and Allegations Number Containment Leak Testing The leaks encountered during the 3 containment integrated leak rate test (CILRT) were numerous and of such magnitude that the CILRT should have been repeated after repairs. 4 Prerequisite Testing The prerequisite testing was being conducted by craft personnel who were not properly qualified; system test engineers (STEs) are signing off tests that were actually con-ducted by craft personnel without the STEs having personally wit-nessed the tests; and, test documentation was being signed by STEs, thereby making it look as though the tests were performed by STEs. 5 Preoperational Testing The preoperational testing was flawed because several system test engineers may work on the same system or one may test a part of many systems, a condition causing confusion and the possibility of omissions; there was a dual number-ing system causing confusion, overlap, and possible omissions; l STEs were not provided with a computer printout informing them ! of all required system tests; cal-culations for the instantaneous
- trip settings for approximately 100 circuit breakers were incor-rectly performed; portions of pre-requisite tests were used to meet Final Safety Analysis Report (FSAR) commitments; system prerequisite and preoperational tests did not always include an energized func-tional test; and STEs were not provided with current design information.
6 Management Attitude TUEC startup management had a
- ' tendency to relax standards when-ever interpretation of commitments or NRC requirements allowed, instead of taking a conservative approach in the interest of public health and safety.
Comanche Peak SSER 7 J-12
7 QA Surveillance of There was minimum surveillance by Testing Activities QA of testing activities. 3.2.2 Test Program Group TRT reviewers were assigned to the Test Program Group based on their techni-cal expertise, capabilities and experience in nuclear power plant operations, testing, QA/QC, inspection program management, and regulatory activities. The Group consisted of a leader from the NRC Region I staff who had previous experience with nuclear power plant testing programs and allegation followup; the NRC Resident Reactor Inspector (operations) who had recently been assigned to CPSES; and two NRC contractor personnel from EG&G, Idaho. In total, the Group represented over 99 years of experience in the nuclear power field. 3.2.3 Findings for Test Program Issues The Test Program Group found the concerns and allegations in Categories 2 (Unit 2 Testing) and 7 (QA Surveillance of Testing Activities) to be without basis. (See Attachment 2, TP Categories 2 and 7.) The allegation in TP Category 6 (Management Attitude) yielded some isolated cases which could have been perceived to be less than conservative and, therefore, was con-sidered to have a valid basis. (See Attachment 2, TP Category 6.) The concerns and allegations in TP Categories 1 (Hot Functional Testing), 4 (Prerequisite Testing), and 5 (Preoperational Testing) were generally found to have valid bases. However, none were found to be of safety significance or, with the exception of one in Test Program Category 1 and one in Test Program Category 5, to have generic implications. In general, the testing activities reviewed by the TRT were carried out in compliance with NRC regulations and FSAR commitments. However, during its review of TP Category 1, the TRT found that three HFT data packages were approved by the TUEC Joint Test Group (JTG) that failed to meet all of the objectives stated in the test procedures. appeared to violate 10 CFR 50, Appendix B, Criterion V. (See Attachment 2,These TP Category 1.) In Category 4, the TRT found that TUEC startup management authorized, by memorandum, test support craftsmen to verify initial conditions for certain prerequisite test procedures in violation of Startup Administrative Procedure CP-SAP-21, " Conduct of Testing." a violation of 10 CFR 50, Appendix B, Criterion V.This In instruction also5,appears TP Category the TRTto be found that system test engineers (STEs) were not on controlled distribution for design changes applicable to systems to which they were assigned; rather, they were required to obtain this information on their own initiative from the document control center prior to starting a test and were then required to incorporate that information, as applicable, into the test procedure. While the TRT did not identify any specific problems as a result of this practice, it considers this practice to be weak since it relies too heavily on the motivations and initiatives of test personnel to ensure that they have current design information when they are developing test procedures and before conduct-ing tests. Typically, these are periods when they could be under more than normal pressure. Additionally, because of the number and nature of the problems found in the document control system by the TRT QA/QC Group, the TRT could not reasonably conclude that the document control system problems identified did not affect testing activities. (See Attachment 2, TP Category 5.) Comanche Peak SSER 7 J-13 i
1 The Group found no safety significance for the allegations in TP Category 3 (Containment Integrated Leak Rate Testing), but concluded that a generic pro-blem could exist because when the CILRT leakage rate was calculated by a method different from that which was committed to in the FSAR, the FSAR had not been amended to reflect that change. The TRT questioned the TUEC procedure for documenting and identifying FSAR deviations to the NRC. The TRT also questioned that method of calculation, which was not endorsed by the NRC. Additionally, the TRT found that the preoperational CILRT was conducted with three isolated electrical penetrations, a condition which did not provide the configuration that the Containment Building would have during normal operation. These items were considered unresolved on the TRT and The werelatter forwarded to the two have since NRC Office of Nuclear Reactor Regulation for action. been resolved to the satisfaction of(See the Attachment NRC as reflected in Item (36) 2, TP Category 3.) in Sec-tion 1.7 of Comanche Peak SSER 6. 3.2.4 Overall Assessment and Conclusions Except for those unresolved issues identified in the foregoing sections, the testing activities included in the TRT review effort were generally found to have been carried out in compliance with NRC regulations and FSAR commitments. Adequate administrative controls had been established for the testing program, and it appeared that they were generally implemented properly. The test engi-neers were appropriately experienced and qualified to conduct and supervise a testing program, and were found by the TRT to be generally dedicated and responsible individuals, which contributed significantly to the success of the program. The startup group personnel, in interviews conducted by the TRT, were found to be candid, knowledgeable, and very responsive to TRT requests for information. The TUEC startup group relies heavily on the accuracy and completeness of the design documents, which are included in the document control system, in its preparation of test procedures and during the conduct of testing. A number of problems were identified in the document control system by the TRT QA/QC Group during its review. While the TRT Test Program Group did not find that these problems adversely affected those portions of the testing program that it included in its review, the TRT cannot conclude with reasonable assurance that the document control system problems had no adverse effect on testing activi-ties. Therefore, the TRT will require TUEC to provide NRC with assurance that all structures, systems, and components were properly and completely tested before it can draw a final conclusion with regard to the testing program. Comanche Peak SSER 7 J-14
- - - -- . . . - - - - . . - - - . - . _ . . . . . - . - = . -_
l J t
- 4. Actions Required of TUEC TUEC shall submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified in the following subsections. This program plan and its implementation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1. The program plan should
} address the root cause of each problem identifi 4 on safety-related systems, programs, or areas. ed and The its generic colles.tive implications significance of these deficiencies should also be addressed. The program plan should also
! include from the proposed occurring in theTUEC future.action to assure that such problems will be precluded in the following sections. The specific actions required of TUEC are described 4.1 Electrical and Instrumentation (E&I) Area 4.1.1 Electrical Cable Terminations '(See Attachment 2 for E&I Category 1) 1 -
! Reevaluate and redisposition all NCRs related to vendor-installed terminal lugs in ITT Gould-Brown Boveri switchgear; and perform and document the results of engineering analyses to justify any resulting "use as-is" dispositions. 1, j - 1 Develop adequate installation and inspection procedures to ensure (1) the
- operability of those circuits which contain butt splices in panels, I (2) that the wire splicing eaterials and methods used are qualified for
< anticipated adjacent to each services conditions, and (3) that splices are not located other.
Reinspect all safety related and associated terminations in the control j room panels and in the termination cabinets in the cable spreading room f to verify design that their documents. locations are accurately depicted on all current Should the results of this reinspection reveal an unacceptable level of nonconformance to design documents, the scope of I this reinspection effort shall be expanded to include all safety-related and associated terminations at CPSES. Clarify procedural requirements and provide additional QC inspector training with respect to the areas in which nuclear heat-shrinkable-1 sleeves are required on splices,'and ensure that (1) such sleeves are installed where required, (2) all QC inspections requiring witnessing for splices have been performed and properly documented, and (3) all butt splices are properly identified on the appropriate design drawings and are physically identified within the appropriate panels. Evaluate the adequacy of the QC inspection program as related to the deficiencies identified above to establish root causes and appropriate i corrective actions. These actions shall be integrated with other actions addressed under QA/QC Category 8, "As Built."
! 4.1.2 Electrical Cable Tray and Conduit Installation (See Attachment 2 for , E&I Category 2)
I I Comanche Peak SSER 7 J-15 i.
- Evaluate the adequacy of craft personnel training in the use of installa-tion manuals to establish root causes and appropriate corrective actions. This action shall be integrated with other actions concerning craft personnel training addressed under QA/QC Category 8, "As Built." 4.1.3 Electrical Equipment Separation (See Attachment 2, E&I Category 3)
- Reinspect all panels at CPSES, in addition to those in the main control room for Units 1 and 2, that contain redundant safety-related cables within conduits or safety and nonsafety-related catfles within conduits, and either correct each violation of the separation criteria, or demonstrate by analysis the acceptability of the conduits as a barrier for each case where the minimum separation is not met.
- Reinspect all panels at CPSES, in addition to those in the main control room identified in Table 1 of SSER for E&I Category 3, and either correct each violation of the separation criteria concerning separate cables and cables within flexible conduits, or demonstrate by analysis the adequacy of the flexible conduit as a barrier.
- Correct two instances of violation of the separation criteria inside panels CPI-EC-PRCB-09 and CPI-EC-PRCB-03 concerning a barrier that had been removed and redundant field wiring not meeting minimum separation.
- Submit the analysis that substantiates the acceptability of the criteria stated in the electrical erection specifications governing the separation between independent conduits and cable trays.
- Evaluate the adequacy of the QC inspection program as related to the deficiencies identified above to establish root causes and appropriate corrective actions. These actions shall be integrated with other actions addressed under E&I Category 6, " Electrical QC Inspector Training and Qualifications," and QA/QC Category 8, "As Built."
4.1.4 Control Room Ceiling Fixture Supports (See Attachment 2, E&I Category 4)
- Substantiate (1) the adequacy of the overall seismic support system installation for all the items located above the ceiling in the control room, including nonsafety-related conduit, suspended ceiling and lighting and (2) the adequacy of the seismic support system installation for nonsafety-related conduit in Seismic Category I areas of the plant other than the control room. This action shall be integrated as appropriate with other actions addressed under Civil / Structural Category 14, " Seismic Design of Control Room Ceiling Elements."
4.1.5 Electrical QC Inspector Training / Qualifications (See Attachment 2, E&I Category 6)
- Evaluate the testing program for QC electrical inspector qualifications and develop a testing progrem which optimizes administrative guidelines, procedural requirements and test flexibility to assure that suitable proficiency is achieved and maintained.
Comanche Peak SSER 7 J-16
Review all the electrical QC inspector training, qualification, certifica-tion, and recertification files against the project requirements as docu-mented in the FSAR and provide the information in such a form that each requirement is clearly shown to have been met by each inspector. If an inspector is found to not meet the training, qualification, certification, or recertification requirements TUEC shall then review the records to determine the adequacy of inspections made by the unqualified individuals and provide a statement on the impact of the deficiencies noted on the safety of the project. - Justify the allowance to administer separate (waiver) tests, as permitted by procedures, in lieu of examinations administered by independent professional eye specialists. These actions shall be integrated, as appropriate, with other actions addressed under QA/QC Category 4, " Training and Qualifications." 4.2 Test Program (TP) Area 4.2.1 Hot Functional Testing (HFT) (See Attachment 2, TP Category 1) Review all completed preoperational test data packages to ensure there are no instances where test objectives were not met, or prerequisite conditions were not satisfied. Address the four items identified by the TRT, along with appropriate resolution. Since the review of data obtained from the deferred preoperational testing is a function of the Station Operations Review Committee (SORC), amend the FSAR to reflect that the 50RC, and not the Joint Test Group (JTG), will perform these reviews. Incorporate the information necessary to provide traceability between thermal expansion test monitoring locations and measuring instruments. Also establish administrative controls to ensure appropriate test and measuring equipment traceability during future testing and plant operations. 4.2.2 Containment Integrated Leak Rate Testing (CILRT) (See Attachment 2, TP Category 3) TUEC has identified deviations from FSAR commitments related to the CILRT. TUEC shall identify all other deviations from FSAR commitments which were not previously identified to NRC. 4.2.3 Prerequisite Testing (See Attachment 2, TP Category 4) Rescind the startup memorandum (STM-83084), which was issued in conflict with CP-SAP-21, and ensure that no other memoranda were issued which are in l conflict with approved procedures. Also, conduct a review of all other prerequisite test recordt to determine those that had prerequisites signed by craf t personnel, and assess the impact of those improperly verified on subsequent testing activities. Comanche Peak SSER 7 J-17
. _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . . _ _ _ _ _ - _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ __m-_ ______.___ _ . _ _ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ .
4.2.4 Preoperational Testing (See Attachment 2, TP Category 5) .
- Establish measures to provide greater assurance that STEs and other respon-sible test personnel are provided with current controlled design documents and change notices.
- Provide NRC with reasonable assurance that the document control system problems identified by the TRT QA/QC Group did not affect the testing activities.
One action required in the enclosure to the NRC letter of September 18, 1984, to TUEC (Attachment 3) was that "TUEC shall evaluate the required plant conditions for the deferred preoperational tests against limiting conditions in the proposed technical specifications and obtain NRC approval where devia-tions from the technical specifications are necessary." This requirement is no longer applicable since the TRT has been informed by TUEC that these tests will be conducted prior to fuel load. TUEC was also required in the September 18, 1984, letter to " justify to NRC the conduct of preoperational CILRT (Type A Test) with penetrations isolated and leakage rate calculation in accordance with ANSI /ANS 56.8 rather than ANSI N45.4-1972" and to " identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rather than ANSI N45.4-1972." These issues have been resolved (see page J-83); accordingly, the actions are no longer required. I Comanche Peak SSER 7 J-18
ATTACHMENT 1 LISTING OF TECHNICAL CONCERNS AND ALLEGATIONS I. Electrical and Instrumentation Allegation Number Chacterization Category Page Number AQE-1 An electrical inspector was pressured not 5 J-49 to write nonconformance reports (NCRs) in several instances. In one case, a QC Supervisor instructed him not to write an NCR for control room cables that were 1 removed without proper documentation. AQE-2 A cable was removed from the Safeguards 5 J-49 Building without proper documentation. An NCR was prepared, but it was uncer-tain whether that NCR was fully gen-erated, processed, and disposed. AQE-3 An inspector was told to close-out an 5 J-49 NCR that described repair of a flex-ible conduit in the Fuel Handling Building when the conduit had been replaced rather than repaired. AQE-4 Unqualified inspectors were told to 5, 6 J-49, J-55 close-out NCRs. AQE-5 An inspector was asked to close-out an 5, 7 J-49, J-59 NCR on a cable tray to allow craft personnel to pull cable. The inspector did not close out the NCR because the nonconfonsing conditions, including trash in the tray, cuts in cable jackets, and interwoven cable, still existed. The supervisor assigned another inspector who closed out the NCR. AQE-6 Electrical inspectors were directed by 3, 8 J-37, J-63 a QC supervisor to violate inspection procedures. AQE-7 A QC supervisor instructed electrical 9 J-67 inspectors not to perform required in-process inspections, but only to inspect completed work. J-19
I. Electrical and Instrumentation (Continued) Allegation Category Page Number Number Characterization Some electrical inspectors were not ade- 6 J-55 , AQE-8 quately qualified, were given help to pass their certification tests, and had incor-rect descriptions of prior electrical or inspection experience on their employment applications. t AQE-9 Field copies of drawings used by electri-cal inspectors to perform inspections were not always the most up-to-date version. (Transferred to the QA/QC Category 2.) Craftsmen installing conduit supports 2 J-33 AQE-10 were not properly trained, thus necessitating extensive rework. QC Supervisors were overly sympathetic 3 J-37 AQE-11 to the needs of production managers. Some electrical terminations were 1,5,6 J-27, J-49, AQE-12 J-55 accepted by inadequately qualified inspectors; these terminations did not conform with the drawings. Terminal lugs of improper size and type 1 J-27 AE-13 were used in certain panels, and impro-per cable splices existed within various panels. Attachments were installed on cable trays 2 J-33 AE-14 and hangers at the 810-ft elevation of the Safeguards Building without required and approved design changes. Installed safety-related cables and con- 3 J-37 AE-15 duits in the reactor control panel in the control room did not conform to separation criteria. A Safeguards I panel at the 790-ft eleva- 1 J-27 AE-16 tion had loose bus bars and ground wire connections. Field run conduit, drywall, and lighting 4 J-45 AE-17 installed above control room pancis were classified nonseismic and inadequately supported. Cables were butt spliced inside panels 1, 8 J-27, J-63 AE-18 in violation of procedures. J-20
I. Electrical and Instrumentation (Continued) Allegation Number Characterization Category Page Number AE-19 Cable trays were overfilled. 7 J-59 AE-20 Separation requirements in the Electri- 3, 8 J-37, J-63 cal Erection Specification for the cable spreading room were inconsistent with the requirements of Regulatory Guide (R.G.) 1.75. The installation of inde-pendent safety-related cable trays and conduit between safety-related and nonsafety-related raceway did not con-form with R.G. 1.75. AQE-21 (This allegation is assessed in Test Programs, Category 5, "Preoperational Test Program.") AE-22 Cable butt splices existed in panels 1, 5 J-27, J-49 without authorization or documentation on drawings. AQE-23 Many requirements were deleted by re- 8 J-63 visions to post-construction electrical inspection procedures. AE-24 A cable tray supported by a temporary 5 J-49 hanger fell, damaging instrumentation cables entering the control room. AQE-25 Electrical QC inspectors were required 5 J-49 to submit draft NCRs to their supervisors for approval in contradiction of site procedures. AE-26 Conductors with two different gauges were 1 J-27 terminated at some lugs, and many termina-tions were loose. AE-27 Loose elbow termination conduit fittings 2, 5 J-33, J-49 were found at the east and south ends of the Unit 1 diesel generators. NCRs were written, but dispositioned use-as-is. Ati-28 Cables were not trained by use of good 7 J-59 workmanship in the Unit 1 cable spreading room and in junction boxes 1058 and 1059. An NCR dispositioned this condition as acceptable because of proper cable bend radii, but the workmanship problem was not addressed. J-21
I. Electrical and Instrumentation (Continued) Allegation Category Page Number Number Characterization Sides were cdded to some cable trays 2, 7 J-33, J-59 AE-29 because the trays were overfilled. There may be density and compaction pro- 7 J-59 AE-30 blems in cable trays with excessive fill. There were instances of inadequate separa- 2 J-33 AE-31 tion between process piping and cables that required that notches be made in insulation and in metal barriers between insulation and cables. Because of complaints from craft 8 J-63 AQE-32 personnel, four revisions were made to QI-QP 11.14-12 that deleted inspection requirements. There were prevalent use-as-is dis- 5 J-49 AQE-33 positions written for NCRs generated with respect to the Electrical Erection Specification. A cable jacket was damaged when a Bisco 5 J-49 AQE-34 Seal was removed using a threaded rod. The resulting NCR was dispositioned use-as-is. 5 J-49 AQE-35 Non-Q fuse blocks were installed where Q blocks were required. The NCR was dis-positioned use-as-is because both types of blocks were ordered under the same material specification. Vendor-installed terminal lugs in General 1, 5 J-27, J-49 AQE-36 Electric motor control centers were excessively bent, and the resulting NCR had not been dispositioned. The dispositions of NCRs involving rework 5 J-49 AQE-37 of terminal blocks were questionable. An individual performed an undocumented 5 J-49 AQE-38 repair to a solenoid. Post-construction inspection procedures 1, 8 J-27, J-63 AQE-39 were revised to delete requirements after numerous loose terminations were found in lighting system terminal boxes. J-22
Z. Electrical and Instrumentation (Continued) Allegation Number Characterization Category Page Number AQE-40 Some NCRs were closed out by stating 5 J-49 that the nonconforming condition was not addressed in the Electrical Erection Specification. AQE-41 An NCR was written because worn lighting 5 J-49 restraint cable crimp gauges were causing indeterminate inspection results. (Also addressed under QA/QC Category 6, "QC Inspection.")* AQE-42 An individual was pressured not to write 5 J-49 NCRs during turnover. AQE-43 Some inspection reports were written 9 J-67 without the reinspection needed to clear cable tray inspection item removal notices. AQE-44 An individual was not satisfied with a 3, 8 J-37, J-63 use-as-is disposition for an NCR involv-ing a cable separation problem in the Fuel Handling Building. AQE-45 There were questionable dispositions for 5 J-49 NCRs involving inadequate thread engage-ment between a conduit fitting and damaged cable. AQE-46 Post-construction inspection procedures 1, 8 J-27, J-63 were revised to delete attributes with frequent problems, such as loose light-ing terminations. AQE-47 Many NCRs were dispositioned use-as-is. 5 J-49 AQE-48 Some NCR evaluations inaccurately de- 5 J-49 scribed workmanship as "not compromised" when it had been poor. AQE-49 Excessive rework was required to achieve 3 J-37 proper separation. QThe TRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER. l l J-23 l _ _
l I. Electrical and Instrumentation (Continued) Allegation Category Page Number Number Characterization Cables in the cable spreading room were 5, 7 J-49, J-59 AE-50 spliced in violation of regulatory requirements. AE-51 A conduit was about 3 feet below a cable 3 J-37 tray in the Control Room Building, perhaps violating separation criteria. Revision 15 to a post-construction 8 J-63 AQE-52 inspection procedure eliminated the re-quirement to inspect large pieces of equipment such as 6.9 kV motors. Separation between two conduits was 3 J-37 AQE-53 accomplished only after improper conduit bending. Ladder type cable trays should not qualify 3 J-37 AQE-54 as barriers; therefore, the 1-inch separa-tion criteria between ladder-type trays and conduits routed under the trays should not apply. AH-14 Attachments were installed on cable trays 2 J-33 and hangers without required design changes. (Also, inadequate spacing of seismic sup-ports for cable trays and material trace-ability for cable tray supports.) SRT-10 The effect of the weight of thermolag 7 J-59 material on cable trays requires evaluation. II. Test Programs Allegation Number Characterization Category Page Number AT-1 The Hot Functional Test was deficient in 1 J-69 that major components and equipment were not installed at the time of testing. j AT-2 Significant modifications have been made 1 J-69 l or planned which invalidate the Hot Functional Test. AT-3 TUEC does not intend to confirm per- 1 J-69 formance of major components and equip-ment until after fuel loading. J-24
~ ^ II. Test Programs (Continued) Allegation Number Characterization Category Page Number AT-4 Neither TUEC nor NRC Region IV staff 1 J-69 noticed that major components and equip-ment were not installed prior to the Hot Functional Test. AT-5 The Hot Functional Test was inadequate 1 J-69 because it did not include accident con-ditions, such as earthquakes and Loss of Coolant Accidents (LOCAs); the deficien-cies found during the Hot Functional Test demonstrate that the plant cannot be operated successfully during an accident. AT-6 The willingness of both the applicant and 1 J-69 the NRC Region IV staff to accept Hot Functional Test results which are defi-cient makes it impossible to rely on the test results to prove CPSES is safe. AT-7 Problems revealed by the Hot Functional 3 J-81 Test, and related containment and leak-rate tests, are so extensive and of such magnitude that they must be corrected before fuel load. AT-8 In order for the health and safety of the 1 J-69 public to be assured, Texas Utilities Electric Company (TUEC) must correct pro-blems in design and construction, following which they must conduct addi-tional tests, including a Hot Functional Test, until such time as the tests can be run successfully with all finalized equip-ment in place. AT-9 Neither the NRC staff nor TUEC has 1 J-69 informed the Atomic Safety and Licensing Board (ASLB) of the extent and magnitude of the problems uncovered in the test program. AT-10 The ASLB itself should closely monitor 1 J-69 the successful completion of tests and reinspections. J-25
II. Test Programs (Continued) Allegation Catcaory Page Number Number Characterization i The ASLB should recognize that test 1 J-69 AT-11 - result evaluations performed by TUEC ! and the NRC staff were incomplete and i inaccurate. The ASLB should consider i these inadequacies when examining testi-many given by TUEC and the NRC staff and when making its decisions. < i Separate tests should be required for 2 J-79 AT-12 Unit 2, rather than relying on tests performed for Unit 1 to reveal problems. The ASLB should order a complete rein- 1 J-69 AT-13 spection of all components, equipment, welding, and "everything" before allow-ing fuel loading. Prerequisite testing was performed by 4 J-85 AT-14 unqualified craft personnel; system test ; engineers (STEs) were signing test docu-ments for tests performed by craft per-sonnel when the STEs were not present; < and test documentation reflected test performance by STEs when tests were act-ually performed by craft personnel. The preoperational test program was flawed 5 J-91 AT-15 > because: (1) there was a dual numbering ! system causing confusion, overlap, and pos-sible omissions; (2) STEs were not provided . with a computer printout informing them of all required system tests; (3) calculations for the instantaneous trip settings for approximately 100 circuit breakers were incorrectly performed; (4) portions of pre-requisite tests were used to meet Final Safety Analysis Report (FSAR) commitments; (5) system prerequisite and preoperational tests did not always include an energized functional test; and (6) STEs were not provided with current design information. l TUEC upper management liberally inter- 6 J-97 AT-16 ; preted their FSAR commitments. There were numerous problems with the 1 J-69 i AT-17 thermal expansion test. l 7 J-99 AT-18 There was minimal QA surveillance of j test program activities. i J-26 : f
, __ .- - = _ , . , - .w,-
ATTACHMENT 2 ASSESSMENT OF INDIVIDUAL TECHNICAL CONCERNS AND ALLEGATIONS IN ELECTRICAL AND INSTRUMENTATION AND TEST PROGRAM AREAS
- 1. Allegation Category: Electrical and Instrumentation 1, Electrical Cable Terminations 2.
Allegation Number: AE-13, AE-16, AE-18, AE-22, AE-26, AQE-12, AQE-36, and parts of AQE-39 and AQE-46.
- 3. Characterization: It is alleged that:
Terminal lugs of improper size and type were utilized on cables in various panels and that improper cable splices existed within certain panels (AE-13). Loose bus bar and ground wire connections existed in a safeguards panel (AE-16). Cables were butt spliced inside panels in violation of procedures (AE-18). Cable butt splices existed in panels without authorization or without being documented on drawings (AE-22). Cable termination connections.were loose and improper-sized lugs were used on cable terminations (AE-26). Cable terminations not in conformance with drawings were accepted by quality control (QC) personnel (AQE-12). Vendor-installed terminal lugs were excessively bent and correspond-ing nonconformance reports (NCRs) were improperly dispositioned (AQE-36). Certain quality assurance / quality control (QA/QC) matters related to cable terminations were improperly implemented. The general concerns expressed in these allegations are within the scope of the above allegations and are addressed below as appropriate (parts of AQE-39 and AQE-46). 4. Assessment of Safety Significance: The implied safety significance of these allegations is that improper installation of butt splices and cable connections, disagreement of the installation with as-built drawings, or J-27
improperly dispositioned NCRs could place the quality of the installation in question. Sample of Safety-Related Termination Installation. Since many of the alleged conditions identified in AE-13, AE-16, AE-26, parts of AQE-39 and AQE-46 were located in equipment containing nonsafety-related cabling, the NRC Technical Review Team (TRT) also sampled safety-related installations to determine whether similar conditions existed within them. Sixteen safety-related items (control panels, annunciator cabinets, termination cabinets, motor control centers, and switchgear) were inspected for the following items:
- Proper size lugs used relative to cable size and screw size (AE-13).
Tightness of bus bar and ground wire connections and terminal lugs on terminal blocks (AE-16).
- General workmanship for such items as shaved lugs, proper washers, and bend radii (AE-26).
1 The TRT found no unacceptable conditions with the terminations inspected, including those associated with AE-13, AE-16, AE-26 and parts of AQE-39 and AQE-46. Butt Splices. Allegations AE-13, AE-18, and AE-22 concerned butt splices in panels that could be in violation of regulatory requirements and site procedures. The practice of butt splicing cables in panels was allowed on a limited basis, as specifed in Section 8.1.5.2.4 of Amendment 44 to the Final Safety Analysis Report (FSAR). The NRC staff reviewed Texas Utili-ties Electric Company's (TUEC's) justification for permitting butt splices inside panels (correspondence from M. Srinivasan, NRC Power Systems Branch to B. J. Youngblood, NRC Licensing Branch, July 30, 1984), and concluded that the practice is acceptable on a limited basis, subject to the fol-lowing conditions: That adequate provisions be included in the installation procedures to verify operability of those circuits for which splices are being used,
- That the wire splices used are qualified for anticipated service conditions, and
- That splices are staggered within the panel so that they are not adjacent to each other in the same wire bundle and pressing against one another.
The TRT inspected butt splices in safety-related panels to determine whether they were installed in accordance with the requirements stated in Texas Utilities Generating Company (TUGCO) procedure QI-QP-11.3-28, Revision 21, " Class 1E Cable Terminations." The TRT also interviewed one alleger to clarify one allegation concerning butt splices. J-28 1
The TRT found the splices to be in conformance with all procedural requirements set forth by TUGC0 which did not include the three conditions for acceptability stated above, which the NRC considers important to assure the adequacy of these splices, with the following exception. All splices inspected were missing the " nuclear heat-shrinkable cable insula-tion sleeves," as required by paragraph 3.2.15 of the procedure for 600-volt control anc instrumentation connections. Due to this recurrent condition, the TRT reviewed the QC inspection reports for 12 butt splices and found the following: Nine of these splices were documented on the inspection form designated in paragraph 3.3 of the procedure for post-installation inspections instead of on the correct form designated for witnessing-type inspections. It should be noted that all splices were required to be witnessed by QC personnel per paragraph 3.1.d of the procedure. Six of the nine incorrect forms contained handwritten notes by the inspector indicating that he had witnessed the splice; however, no reference was added to indicate that the installation of the heat-shrinkable sleeves was required to be witnessed. The remaining three of the nine incorrect forms did not indicate that the splices had been witnessed. For three splices which were documented on the correct forms, the forms all contained an "N/A" (not applicable) handwritten by the inspector on the line indicating that the installation of the heat-shrinkable sleeve was witnessed. In summary, the lack of awareness of where the heat-shrinkable sleeves should be installed, as reflected in the QC inspection form, when the high percentage of missed and/or improperly documented inspections requiring witnessing, indicated that craft and inspection personnel lacked familiarity with these procedural requirements. This apparent lack of familiarity may be indicative of poor training. (See Electrical and Instrumentation Category 6, " Electrical QC Inspector Training / Qualification.") Nonconformance of Cable Terminations with Drawings. Allegation AQE-12 involves QC inspectors " buying off" terminations that did not conform to drawing requirements. In view of the lack of specific information con-cerning this allegation, the TRT selected 380 cables, involving 1600 individual terminations, and inspected them in detail with respect to drawing requirements. This inspection revealed that six cables (five of which are safety related) were not terminated in accordance with current drawings. These six cables are: (1) E0139880 in panel CP1-ECPRCB-14, (2) E0110040 in panel CPI-ECPRTC-16, (3) E0118262 in panel CP1-ECPRTC-16, (4) NK139853 in panel CP1-ECPRCB-02 (non-safety), (5) EG104796 in panel CP1-ECPRTC-27, and (6) EG021856 in panel CPX-ECPRCV-01. J-29 l
Terminal Lugs. Allegation AQE-36 involved vendor-installed Amp Product Corporation (APC) terminal lugs in ITT Gould-Brown Boveri, 6.9 kV switch-gear being excessively bent in the area between the ring and the barrel. The TRT discovered 16 NCRs (E-84-01066 through E-84-01081) issued early in April 1984, which documented this condition. The TRT review of TUEC action taken regarding these NCRs revealed the following:
- The NCRs described the APC lugs either as being bent in excess of 60 degrees or twisted.
- The documented record of a telephone conversation between TUEC and the representative of the lug manufacturer (reference letter VBR-16624) stated that lugs bent to 90 degrees one time were to be considered acceptable; that lugs bent to 120 degrees could be accept-able after utilizing an engineering evaluation by the end-user; and that although lugs bent to 120 degrees would not maintain their full mechanical strength, they would maintain their electrical character-istics. This acceptance criteria for field bent lugs was changed by APC due to the dispositioning of NCR E-84-00972 regarding the General Electric (GE) motor control center (MCC) thermal overload relay replacement program.
The TRT findings regarding the disposition of these flCRs were as follows:
- The disposition block of the NCR, form stated that many of the lugs were " determined not to pose an equipment serviceability problem."
However, there was no reference to or evidence of an engineering evaluation, as required by the lug manufacturer prior to a change in the acceptance criteria on NCR E-84-00972.
- Only the " bent" condition of the lugs was addressed by both the vendor representative and TUEC engineering. Neither the mechanical strength nor the electrical characteristics were ever addressed with respect to " twisted" lugs.
The TRT determined that these NCRs were improperly dispositioned in that the full scope of the identified problem was not addressed and the "use-as-is" dispositions were not adequately justified.
- 5. Conclusions and Staff Positions: The TRT concludes that concerns exist in the following areas relative to cable terminations:
- The adequacy of butt splices in safety-related panels concerning operability, qualification for service conditions, and relative loca- .
tion of splices to each other (AE-13, AE-18 and AE-22).
- The acceptability of vendor-installed terminal lugs in ITT Gould-Brown Boveri switchgear (AQE-36).
- Safety-related terminations which are not in conformance with current drawings (AQE-12).
- The adequacy of QC inspections and supporting documentation, parti-cularly with respect to termination activities requiring witnessing by QC personnel.
J-30
- 6. Action Required:
fuel loaa: TUEC shall accomplish the following actions prior to (a) Reevaluate and redisposition all NCRs related to vendor-installed terminal lugs in ITT Gould-Brown Boveri switchgear, taking into con-sideration the effects of twisted as well as bent lugs, and perform and document the results of engineering analyses to justify any resulting "use as-is" dispositions. (b) Develop adequate installation and inspection procedures to reinspect all existing butt splices to ensure (1) the operability of those circuits which contain butt splices in panels, (2) that the wire splicing materials and methods used are qualified for anticipated service conditions, and (3) that splices are staggered within the panel so that they are not adjacent to each other in the same bundle. (c) Reinspect all safety-related and associated terminations in the con-trol room and in the termination cabinets in the cable spreading room to verify that their locations are in accordance with all current design documents. Should the results of this reinspection reveal an unacceptable level of nonconformance to design docum nts, the scope of this reinspection effort shall be expanded to include all safety-related and associated terminations at Comanche Peak Steam Electric Station (CPSES). (d) Provide additional QC inspector training with respect to the areas in which nuclear heat-shrinkable sleeves are required on splices and ensure that (1) such sleeves are installed where required, (2).all QC inspections requiring witnessing for splices have been performed and properly documented, and (3) all butt splices are properly identified on the appropriate design drawings and are physically identified within the appropriate panels. (e) Evaluate the adequacy of the QC inspection program as related to the deficiencies priate identified corrective above to establish root causes and appro-actions. These actions shall be integrated with other actions addressed under QA/QC Category 8, "As Built."* QTRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER. J-31
1, Allegation; Category; . Electrical and Instrumentation 2, Electrical Cable Tray and Conduit Installation 4
- 2. Allegation Number: AQE-10, AE-14, AE-27, AE-29, AE-31 and AH-14.
- 3. Characterization: It is alleged that, in general, there were problems a
with: i Cable tray support installation without required design changes (AH-14).
- Addition of higher sides to cable trays due to overfill (AE-29).
Inadequate clearance of process pipes from cables in cable trays (Q&A specification 2323-ES-100, not met) (AE-31). i Loose concuit fittings (AQE-27). The adequacy of training of personnel installing conduit supports (AQE-10). } The cable tray attachments (clamps) to the seismic supports without
; approved design changes (AE-14, AH-14).
1
- Inadequate spacing of the seismic supports for cable trays (AH-14).
Inadequate material traceability for cable tray supports (AH-14).
- 4. Assessment of Safety Significance: The implied safety significance of these allegations is that the quality of the installation of cable trays and their supports or conduit fittings could be in question.
1 The NRC Technical Review Team (TRT) determined that the first two concerns (AH-14 and AE-29) related to whether the positions of Regulatory Guide 1.29,
" Seismic Design Classification," as augmented by Final Safety Analysis
- Report (FSAR) Section 3.2, were considered by the Texas Utilities Electric Company (TVEC) during design of the support systems for both safety related and nonsafety related cable trays.
The TRT examined cable tray support installation notes and detail drawings, design change authorizations (DCAs),' work packages, physical configuration drawings, and other documents pertinent to its sampling of 29 supports in the Safeguards, Auxiliary, and Control Buildings. The TRT found no deviations from the acceptable criteria for the installation of supports. Welds were not included in this examination; the inspection of
; electrical cable tray support welds is addressed under QA/QC Category 8, "As Built." The TRT also evaluated a sample of cable trays in the cable spreading room to assess the concern about the higher cable tray sides. ~
This evaluation and its conclusions are presented in Electrical Instru- _; , mentation Category 7, " Electrical Cable Installation." The third concern (AE-31) related to process pipe-to-cable-tray clearances not meeting the Gibbs & Hill (G&H) electrical specification 2323-ES-100, as amended by DCA 13045 and DCA 15917. The TRT conducted a walkdown inspection of approximately 2500 feet of cable tray in the auxiliary building and identified 16 cases that appeared not to meet installation guidelines set forth by the acceptable specification above. However, after an examination of the DCAs pertaining to each of the 16 cases, the TRT determined that the DCAs will satisfactorily correct deviations from specifications in the installations for all 16 cases. The fourth concern (AE-27) was the "use as-is" disposition on a non-conformance report (NCR) which reported two loose conduit elbow fittings J-33 l _ , _ . - - . _ - -_-_ -~ - - - - - ~ - - ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
i on the south and east end of the Unit 1 diesel generator. The TRT inspected However, the Unit 1 diesel generator conduit and found two loose fittings. the TRT determined that in the unlikely event of failure of the cables in the loose fittings, the functional capability of the diesel generators would not be affected because those cables were not important in the operation of the diesels. , The fifth concern (AQE-10) was the lack of training of personnel installing conduit supports. The TRT interviewed craft personnel, craft supervisors, and training personnel to determine the availability and effectiveness of the training program, and found that there was a training program for newly This training program hired personnel or transfers into the installation. The interviews revealed l included periodic briefings on procedure changes. l that the training program was not effective because 7 of the 11 crew , members interviewed were not cognizant of Manual 2323-S-0910, " Conduit and ! Junction Box Supports," which is the primary reference manual for installa-tion of supports. Although these seven crew members indicated that they had no need to use this manual in their job assignments, the TRT could not substantiate this assertion. Hence, the lack of awareness of this proce- l dure by craft personnel may be indicative of poor training in the area of procedural requirements. Similar findings in other construction crafts are addressed under QA/QC Category 8, "As Built." The sixth concern (AE-14, AH-14) was that cable tray attachments (clamps) , to the seismic supports were nat installed according to design. The TRT J I inspected 60 cable tray attachments in the Safeguards Building and found no unacceptable cable tray attachments in the sample. The seventh and eighth concerns (AH-14) were that the designed spacing of the seismic cable tray supports was not adhered to during construction and that the supports did not have proper material traceability. The TRT conducted a walkdown inspection of seismic ceble tray supports in the Safeguards and Auxiliary Buildings, and compafed the installed cable tray support spacing with the designed support spacing, including identifica-tion of material traceability for the supports. Two deviations in support spacing were located out of 40 examples inspected. The TRT asked TUEC engineering to provide the analyses for these two deviations because they were outside the designed support spacing. A review of TUEC's documenta-tion of analyses indicated that the spacing maintained was adequate to meet regulatory requirements. Based on the review of engineering
- 5. Conclusion and Staff Positions:
drawings and direct inspection of the installation, the TRT found no indication of construction which was contrary to commitments made in the FSAR Section 3.2, as related to AH-14, AE-29. The TRT determined that DCAs 13045 and 15917 will satisfactorily correct the process pipe-to-cable tray clearance deviations.from specification 2323-ES-100 for every case identified during the walkdown inspection (AE-31). The TRT found no problems with cable tray attachments (clamps) to seismic supports (AH-14, AE-14). The TRT determined that the cable tray support spacing meets design requirements and has proper identification for material traceabil-ity, except for two deviations concerning support spacing which were previously analyzed by TUEC.With regard to loose conduit fittings (AQE-27), the by TRT (AH-14). J-34
concludes that the deficiency raised by the NCR has no safety significance; therefore, the NCR was properly dispositioned. The TRT concludes that the concern highlighted by AQE-10 may be indicative of poor training in the area of procedural requirements.
- 6. Action Required:
action: Prior to fuel load TUEC shall accomplish the following Evaluate the adequacy of craft personnel training in the use of installa-tion manuals to establish root causes and appropriate corrective actions. This action shall be integrated with other actions concerning craft personnel training addressed under QA/QC Category 8, "As Built."* RThe TRT evaluation of QA/QC allegations is in progress and will be published j in a subsequent supplement to this SSER. I J-35
- 1. Allegation Category: Electrical and Instrumentation 3, Electrical Equipment Separation
- 2. Allegation Number: AQE-6, AQE-11, AE-15, AE-20, AQE-49, AE-51, AE-53, AQE-54 and Part of AQE-44.
- 3. Characterization: It is alleged that:
Installation of safety-related cables and conduits inside the reactor control panels in the main control room did not conform to the cable separation criteria (AE-15). Separation between independent safety related cable trays and con-duits, and between them and nonsafety-related trays and conduits in the cable spreading room did not conform to the positions set forth in Regulatory Guide (RG) 1.75, " Physical Independence of Electric Systems." It is also alleged that the separation requirements in Gibbs & Hill (G&H) specification 2323-ES-100, " Electrical Erection Specification," applicable to the cable installation in the cable spreading room, was inconsistent with the separation criteria in the Institute of Electrical and Electronics Engineers (IEEE) Standard 384-1974, "IEEE Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits," as augmented by RG 1.75 (AE-20). Ladder type cable trays did not qualify as acceptable barriers; therefore, the 1-inch minimum separation criteria between separate trays and conduits routed under the trays are not applicable (AQE-54). Nonconformance Report (NCR) E-84-007095 was dispositioned to achieve the required separation between conduits ESB1-4 and C14K30975 without approved conduit bending equipment (AE-53). Post-construction inspection of electrical equipment and raceways in the Fuel Handling Building concerning a cable separation problem was dispositioned "use-as-is" (part of AQE-44). Conduit 22G06343, located in the Control Room Building at the 854-foot elevation, was about 3 feet below cable tray T130CCP38 and thought to violate separation criteria (AE-51). Inspection of the separation of cables did not follow established procedures (AQE-6); quality control (QC) inspection acceptability regarding separation of equipment may have been compromised to meet the needs of production management (AQE-11); and, in numerous cases rework was done to obtain proper separation (AQE-49). These three allegations, in very general terms, raise concerns with cable separation, but do not specifically identify the location of problem areas in the facility. The following discussion will focus on the specific installation concerns of cable separation raised by the allegations. However, the concerns highlighted by AQE-6, AQE-11 and AQE-49 were pursued during the review and inspection of cable separation installations in the various areas of the plant inspected. J-37
Therefore, the NRC Technical Review Team (TRT) findings concerning cable separation disposed of both specific and general concerns raised by the allegations. Quality assurance / quality control (QA/QC) matters raised by allegation AQE-6 are addressed under Electrical and Instrumentation Category 8, " Electrical Procedures." l l
- 4. Assessment of Safety Significance: The implied safety significance of ,
these allegations is that lack of separation may result in a loss of redundancy during design basis accidents and a loss of capability to l mitigate the consequences of accidents or to achieve safe shutdown. Control Room Panels (AE-15, AE-20, AQE-6, AQE-11, and AQE-49). The cri-teria governing the separation of cables inside panels are stated in Section 5.6.2 of IEEE Standard 384-1974, which is endorsed by RG 1.75. Sections 7.1.2.2 and 8.3.1.4 of the Final Safety Analysis Report (FSAR) commit Texas Utility Electric Company (TUEC) to these criteria. l Section 5.6.2 of IEEE Standard 384 states, in part, that the minimum separation distance between redundant Class 1E equipment and wiring internal to the control switchboards (panels) can be established by analysis of the proposed installation. Where the control switchboard materials are flame retardant and analysis is not performed, the minimum i
' separation distance shall be 6 inches. In the event these separation distances are not maintained, barriers shall be installed between re-dundant Class 1E equipment and wiring. The criterion specifying a 1-inch separation between redundant conduits which are considered enclosed raceways is stated in Section 5.1.3 of IEEE Standard 384.
The TRT examined the electrical erection specifications, cable and raceway separation engineering drawings, design change authorizations (DCAs), work packages, and other documents pertinent to the separation of cables, conduits, and devices inside the main control room' panels. The TRT also inspected cables, flexible conduits, terminations, and devices inside six safety-related panels to determine that this equipment was installed in accordance with established separation requirements. In addition, the TRT inspected the separation of cable trays and rigid conduits entering the bottom of the panels from the cable spreading room. The TRT found that the minimum 6-inch air gap or fire retardant barrier between redundant Class 1E panel-mounted devices (including their cable or wire connections) and nonsafety related devices and their connections was maintained in all six panels inspected, except for an instance where a fire-retardant barrier had been removed. The devices involved were FI-2456A, PI-2453A, PI-2475A, and IT-2450, associated with train A, and FI-2457A, PI-2454A, PI-2476A, and IT-2451, associated wi'th train B. These devices were located in auxiliary feedwater panel CP1-EC-PRCB-09. The TRT also found (in panel cpl-EC-PRCB-03, adjacent to the six panels inspected) another instance of redundant safety-related field wiring not being separated by either the 6-inch minimum distance or by a barrier. The field wiring was associated with devices HS-5423 (train B) and HS-5574 (nonsafety related). l J-38
The TRT found no deficiencies in the separation of vertical cable trays and rigid conduits entering the bottom of the panels in the control room floor. The TRT found several instances where (1) redundant safety-related flexi-ble conduits inside the panels were in direct contact with each other and (2) safety and nonsafety related flexible conduits inside the panels were in direct contact with each other. The TRT also found various cases where safety and nonsafety-related cables were in direct contact with safety-related train cables inside thewithin panels.flexible conduits associated with the other redundant These are identified in Table 1. Table 1 Safety or Nonsafety-Related Cables in Contact with Other Safety-Related Conduits in Control Room Panels
- 1. Control Panel CP1-EC-PRCB-02: Containment Spray System Cable No. Train Related Instrument EG139373 8 (green) Undetermined E0139010 A (orange) Undetermined
- 2. Control Panel CP1-EC-PRCB-07: Reactor Control Cable No. Train Related Instrument EG139383 8 (green) Reactor manual trip switch E0139311 A (orange) Undetermined E0139310 A (orange) Undetermined EG139348 8 (green) Undetermined
- 3. Control Panel CP1-EC-PRCB-06: Chemical and Volume Control System Cable No. Train Related Instrument EG139335 B (green) LCV-112C E0139301 A (orange) Undetermined E0139305 A (orange) LCV-1128 NK139605 Nonsafety CSALB-6AB (in bundle)
- 4. Control Panel CP1-EC-PRCB-09: Auxiliary Feedwater Control System Cable No. Train Related Instrument E0139753 A (orange) FK-2453A E0139754 A (orange) FK-2453B EG139756 B (green) FK-2454A '
EG139288 B (green) FK-24548 I EG145780 B (green) FK-2454A B (green) EG145781 FK-2460A A0138622 A (orange Assoc.) HS-2452G-H NK139647 Nonsafety HS-2383 I l l J-39 l
i Table 1, continued
- 5. Control Panel CP1-EC-PRCB-08: Feedwater Control i
Cable No. Train Related Instrument ll B (green) PK-2324 EG140309
- EG139757 8 (green) PK-2328
' NK13957 Nonsafety HS-211A ' ! The TRT discussed with TUEC and G&H representatives the apparent violation of the required 1-inch separation between separate flexible conduits and 6-inch separation between separate cables and cables within flexible con-duits inside the panels. TUEC and G&H representatives indicated that redundant flexible conduits in contact with each other were permitted, ! ' as indicated in the cable and raceway separation typical details drawings, but cables in contact with. cables within flexible conduit were not per-mitted. However, the TRT brought to the attention of the TUEC and G&H i representatives that this type of conduit installation is permitted by Section 5.6.2 of IEEE Standard 384 if such installation can be substan- c i tiated by. analysis. The TRT considered the apparent discrepancies !
described above to be a deviation from the engineering drawings and inconsistent with regulatory requirements. I 3 l Cable Spreadina Room (AE-20, AQE-6, AQE-11, and AQE-49). The criteria 4 governing the separation of redundant safety-related cable trays and con-duits in the cable spreading room appear in Section 5.1.3 of IEEE Standard 384-1974, as augmented by RG 1.75. IEEE Standard 384 states, in part, l j that the minimum separation distance between redundant Class IE cable trays in the cable spreading area can be determined by analysis of the t l l proposed cable installation or, where the conditions of Section 5.1.1.3 ! { (which defines an acceptable tray system) are met, there shall be 1 foot between trays separated horizontally and 3. feet between trays separated , j vertically. Where the minimum separation distance cannot be met, the redundant circuits shall be run in enclosed raceways that qualify as J j barriers, or other barriers shall be provided between redundant circuits.
' The minimum distance between these redundant enclosed raceways and between barriers and raceways shall be 1 inch. ,
The TRT compared these criteria to the requirements set forth in G&H l electrical erection specifications and engineering drawings, concerning l cable tray and conduit separation in the cable spreading room, and identified no deviations. ] The TRT also examined DCAs, work packages, and other doct.ments pertinent to j this issue. In addition, the TRT directly inspected the installation of numerous cable raceways and five termination cabinets in the cable spread-ing room. The TRT found no deviations from separation requirements in the-i J' cable raceways and termination cabinets inspected. Fuel Handlina Buildina Area (Part of AQE-44, AQE-6, AQE-11 and AQE-49). The TRT inspected the cable separation installation in the Fuel Handling l Building area and found that most of the cable trays and conduits ~were designated as nonsafety related. The only safety-related electrical equipment installation in the fuel building area that needed to satisfy i J-40
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- a wpe pe. -*eg-gy. ppg- -p--g ti1t-shtTt'^itP--* QuP *W "4--em-Org ! 1e 'r 3 4-*t*v e 9 '* EPD*'F-T5e**FF * - 7 WgCNe-'-
l separation requirements was associated with the spent fuel system. The TRT found that redundant spent fuel system equipment was located in separate adjacent rooms, except for a common control panel. After examin-ing the separation of cable raceways in the fuel building area and termina-tions and the cables, wires, and devices inside the common control panel, the TRT found no deviations from separation requirements. Potential Harsh Environment Areas (AQE-6, AQE-11, and AQE-49). The TRT examined cable separation installations in those areas of the plant where a high-energy line break could compromise the independence of redundant safety-related equipment. TUEC's damage study group performed studies to determine the need to protect equipment, including cable raceways, that could be affected by a high energy line break. Jet shields were installed to protect safety-related raceways, as required. In the areas where the installation of jet shields was not possible, the affected cable raceways were to be rerouted. The TRT inspected two typical jet shield installations located in the chemical and volume control system (CVCS) piping and valve area and steam generator blowdown area and found that cable separation in these two areas was in accordance with IEEE Standard 384-1974, as augmented by RG 1.75. Remote Shutdown and Transfer Switch Panel Areas (AE-15, AE-53, AQE-6, AQE-11, and AQE-49). The TRT reviewed engineering drawings and electrical erection specifications pertinent to the separation of the safety-related equipment located inside the remote shutdown and' transfer switch panels. The TRT also inspected the cables, wires, and devices (including their cables and wire connections) inside these two panels and the cables enter-ing the top of the panels to determine whether this equipment was installed in accordance with established separation requirements. The TRT found no deviations from separation requirements in these two panels. NCR E-84-007095 concerns the separation between two specific conduits (AE-53) located in the Unit 1 safeguard area, which was established by bending the conduits with unapproved bending equipment. The TRT deter-mined that both conduits dispositioned "use as is" in the NCR were non-safety related and concurred with the "use as is" disposition. Electrical Erection Specification for Separation Criteria (AQE-6, AQE-54, and AE-51). The criteria set forth in IEEE Standard 384-1974, as aug-mented by RG 1.75 and Sections 7.1.2.2 and 8.3.1.4 of the FSAR, have been expressed in specific terms in G&H specification 2323 ES-100, " Electrical Erection Specification." It is alleged that the requirements set forth in this specification governing the separation between independent trays and rigid conduits is inconsistent with the criteria stated in IEEE Standard 384-1974, as augmented by RG 1.75, particularly when ladder type trays and conduits were used as barriers to maintain a 1-inch minimum separation between separate trays and conduits routed under the trays. During its assessment of this allegation, the TRT found a requirement in the electrical erection specification that permitted nonsafety-related rigid conduits to have a minimum separation of 1 inch from the top of open safety-related trays. These requirements appear to be inconsistent with the aforementioned standard and guide. J-41
The TRT determined that no information was included in the FSAR that supported the 1-inch separation between trays and conduits, which is at variance with the requirements of IEEE Standard 384-1974 and RG 1.75. However, the TRT reviewed an existing G&H analysis, including test results, which was used to establish the requirements set forth in specification 2323-ES-100 for a separation of 1 inch between conduits and trays (G&H memorandum EE-863, January 17, 1984, " Cable Tray Conduit Separations"). In essence, the analysis concluded that rigid conduits constituted an acceptable barrier by themselves between the cables inside the conduit and cables inside ladder or open-type trays. Based on the review of electrical specifications, engineering drawings and analyses, inspection reports, procedures, and other pertinent docu-ments and on direct inspection of the installation of cables, conduits, cable trays, terminations and panels in the main control room, cable spreading room, Fuel Handling Building area, potential harsh environment areas, and remote shutdown and transfer switch panel areas, the TRT deter-mined that in general the requirements set forth in IEEE Standard 384, as augmented by RG 1.75 and Chapters 7 and 8 of the FSAR, were satisfied in the areas inspected, except for the following items:
- The TRT could find no evidence that an analysis was performed to support the practice that allowed certain separate safety- and nonsafety-related flexible conduits inside control room panels to be in direct contact with each other or to be separated by less than 1 inch, as required by Section 5.6.2 of IEEE Standard 384 (AE-15).
- The TRT determined that the installation of certain safety- or nonsafety-related cables inside control room panels, which were in direct contact with safety-related flexible conduits associated with the other redundant trains (see Table 1), was inconsistent with engineering drawings and regulatory requirements (AE-15 and AQE-6).
Because acceptability of the flexible conduit as a barrier was not established by analysis, as required by Section 5.6.2 of IEEE Standard 384, the cables must be separated from the conduits inside the panels by a minimum distance of 6 inches, as required by Section 5.6.2 of IEEE Standard 384. (AE-15)
- The TRT determined that the missing barrier (used to separate redun-dant devices in auxiliary feedwater panel CP1-EC-PRCB-09) and the field wiring not being separated by the required 6 inches (inside
- panel CP1-EC-PRCB-03) were the only two instances of Class 1E panel-mounted devices in violation of the separation criteria which require
< corrective action. (AE-15)
- The TRT found no evidence that the existing G&H analysis for estab-lishing the criteria for a 1-inch separation between rigid conduits and cable trays, as stated in G&H Electrical Erection specification 2323-ES-100, had been evaluated by the NRC staff for Comanche Peak.
This analysis should have been referenced in the FSAR. (AE-20)
- 5. Conclusions and Staff Positions: The TRT concludes that the installations reviewed, in general, meet established separation requirements, except for certain safety- and nonsafety-related cables and flexible conduits inside J-42 ,
I l l - ~. _
control room panels which did not meet minimum separation requirements (AE-15). The TRT found no evidence that the lack of separation was justified by analysis. The TRT also concludes that in the absence of analysis to support the lack of minimum separation between separate flexible conduits inside the main control room panels, the existing design arrangement is in violation of regulatory requirements. Furthermore, the lack of separation in the installation of certain cables and flexible conduits is also inconsistent with TUEC's engineering drawings and docu-ments (AQE-6). The lack of analysis to substantiate the adequacy of separation in the above cases may be an indication of weakness in the QA/QC program concerning design control. Category 1, " Design Process."* This area is addressed in QA/QC The TRT concludes that the unjustified installation of cables and flex-ible conduits inside panels that do not meet minimum separation require-ments has potential generic implications. (AE-15) In regard to the criteria for 1-inch separation between rigid conduits and cable trays stated in G&H specification 2323-ES-100 (AQE-54 and AE-51), the TRT concludes that the analyses performed by G&H to support acceptabil-ity of these criteria require NRC evaluation. The present FSAR contains no reference to this analysis. (AE-20) The TRT also concludes that the missing barrier in the auxiliary feedwater panel and the field wiring not being separated by the required 6 inches are two isolated instances of nonconformance and do not have generic implications. The TRT findings on cable separation may be indicative of poor QC per-sonnel inspection.training in procedural requirements for installation and This subject is further addrcssed under Electrical and Instrumentation cations." Similar Category 6, " Electrical QC Inspector Training and Qualifi-findings in other installations are addressed under QA/QC Category 8, "As Built."
- 6. Action Required:
fuel load: TUEC shall accomplish the following actions prior to (a) Reinspect all panels at Comanche Peak Steam Electric Station, in addition to those in the main control room for Units 1 and 2, that contain (1) redundant safety-related conduits, or (2) safety- and nonsafety-related conduits. TUEC shall either correct each violation of the separation criteria or demonstrate by analysis the accept-ability of the conduit as a barrier for each case where the minimum 1 separation is not met. This analysis shall be accomplished in I accordance with the requirements specified in Section 5.6.2 of IEEE Standard 384-1974. Furthermore, in the event that the acceptability of the conduit as a barrier cannot be demonstrated, TUEC shall correct the engineering drawings and related documents to indicate the revised minimum separation of conduits inside the panel for each case. QThe TRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER. '
)
J-43
(b) Either correct each of the violations of separation criteria con-cerning separate cables and cables within flexible conduits found in contact with each other inside main control room panels (Table 1) or demonstrate by analysis the adequacy of the flexible conduit as a barrier. TUEC shall also reinspect all remaining panels in the con-trol room and other areas of the plant containing separate cables and cables within flexible conduit and shall take the same corrective actions as those outlined in Table 1. This analysis shall be accomplished in accordance with Section 5.6.2 of IEEE Standard 384-1974. In the event that the acceptability of the conduit as a barrier cannot be demonstrated, TUEC shall separate cables and cables within flexible conduits by a minimum distance of 6 inches, as required by Section 5.6.2 of IEEE Standard 384. Fur-thermore, TUEC shall correct all appropriate drawings and documents to indicate the revised minimum separation. (c) Take corrective measures to provide a barrier in auxiliary feedwater panel CP1-EC-PRCB-09 separating redundant flow and pressure instruments. (d) Take corrective action to ensure that the required minimum separation of the redundant field wiring identified inside penel CP1-EC-PRCB-03 is maintained either by distance or by an acceptable barrier. (e) Submit to the NRC the analysis substantiating the acceptability of the criteria stated in G&H electrical erection specification govern-This ing the separation between separate conduits and cable trays. analysis shall be supported with the necessary documentation in sufficient detail to perform an independent evaluation of how these criteria were established based on the analysis. (f) Evaluate the adequacy of the QA/QC program as related to the deft-ciencies identified above to establish root causes and appropriate corrective actions. These actions shall be integrated with other actions addressed under Electrical and Instrumentation Category 6,
" Electrical QC Inspector Training and Qualifications," QA/QC Category 8, "As Built," and QA/QC Category 1, " Design Process."
J-44
- 1. Allegation Category: Electrical and Instrumentation 4, Control Room I Ceiling Fixture Supports i
- 2. Allegation Number: AE-17 3.
Characterization: It is alleged that the field run conduit, drywall, and lighting installed in the area above the equipment panels in the control room were classified as nonseismic, and as such were only supported by wires.
- 4. Assessment of Significance: The implied safety significance is that the seismic qualification of certain equipment located above the ceiling in the control room could be indeterminate and consequently its behavior during a seismic event could not be predicted.
The central concern of this allegation is whether Texas Utilities Electric Company (TVEC) considered the positions of Regulatory Guide (RG) 1.29,
" Seismic Design Classification," as augmented by Final Safety Analysis Report (FSAR) Section 3.2.1.2, " Seismic Category II," during the design of the support systems in the control room for the nonsafety-related field run conduit, for the suspended drywall ceiling, and for the lighting fixtures.
Regulatory Guide 1.29 states that nonsafety related structures, systems, or components whose failures could reduce the functioning of any plant feature to an unacceptable safety level or could result in incapacitating injury to occupants of the control room should be designed and constructed so that the safe shutdown earthquake (SSE) would not cause such failure. FSAR Section 3.2.1.2 provides TUEC's commitments to these positions, and designates as seismic Category II the nonsafety-related equipment that will be encompassed by the positions of RG 1.29. Field Run Conduit. The NRC Technical Review Team (TRT) examined conduit seismic installation notes and detail drawings, design change author-izations (DCAs), work packages, physical configuration drawings and other documents pertinent to this issue. The TRT also inspected conduit installation in the area above the control room ceiling and determined that the safety-related conduit was fastened by seismic Category I supports typical of those used in other areas of the facility. The nonsafety-related conduit was secured by supports which were of a dif-ferent design than those for safety-related conduit. None of the non-safety-related conduits examined by the TRT were greater than 2 inches in diameter. In addition, they were not supported by seismic C-ategory I supports and did not have seismic Category II cable restraints. The TRT determined that engineering drawing 2323-5-0910, " Conduit and Junction Box Supports," did not require seismic Category II cable restraints for nonsafety related conduits less than or equal to 2 inches in diameter, but required them for conduits greater than 2 inches in diameter. The TRT also examined similar nonsafety-related conduit installations in other seismic Category I areas of Unit 1 and found that seismic Category II stainless steel cable restraints were used as backup to the nonseismic dead weight supports for the conduits greater than 2 inches J-45
i l l in diameter. The TRT staff also found that the installation of nonsafety- i related conduit less than or equal to 2 inches in diameter in the control room was consistent with that used throughout the plant. Suspended Drywall Ceiling. The TRT found that the suspended ceiling above the central part of the control room was made of drywall sheets These drywall sheets arranged to form a sloping wall around that area. were fastened to a metal framework (metal batten) supported by thin-walled channels (1-1/2-inch by 1/2-inch) attached to the primary building concrete. The metal framework was also attached to the concrete by a system of 1/8-inch stainless steel cables such that if the thin-walled channel supports failed during a seismic event, the weight of the framing and drywall would be assumed by the cabling. Lighting Fixtures. The TRT reviewed the installation'of the lighting fixtures over the control panels and central part of the control room and found that they were supported from an intermediate substructure of "unistrut" by light-weight conduit. The substructure was likewise supported by light-weight conduit from the primary building ceiling. The conduit used is typical of that supporting the light fixtures in 4 suspended ceiling applications. Parallel with each lighting support conduit are two 1/8-inch stainless steel cables which would assumeOther the load if the support conduit or its attachment were to fail. j individual light and reflector assembly fixtures, separate from those supported by the intermediate "unistrut" substructure, were secured by a similar type of conduit and backup cable design arrangement with the cable attached to the edge of the light reflector assembly. 4 I Based on the review of engineering drawings and direct inspection of the installation, the TRT determined that the positions of RG 1.29, as augmented by FSAR Section 3.2.1.2, were not met by the installation of the fixtures located in the area above the panels and central part of the control room. j j As discussed above, the nonsafety-related conduit in the area above the control room suspended ceiling was not fastened by seismic Category I l l supports and/or seismic Category II cable restraints. With regard to the i suspended drywall ceiling, it appeared that the installation met TUEC l commitments to the positions of RG 1.29. However, the final resolu- l tion of this technical issue, including the nonsafety-related conduit ; support system, will depend on the review and approval by the TRT of an i analysis to be provided by TUEC concerning the adequacy of the seismic j support system installation in the control room. i The TRT inspected selected seismic Category I areas of the plant, reviewed associated engineering drawings, and determined that only nonsafety-related i conduits of less than or equal to 2 inches in diameter were not fastened by seismic Category II cable restraints. ! J
- 5. Conclusions and Staff Positions: The TRT concludes that the installation of the nonsafety-related conduit in the control room appears to be inconsistent with the positions of RG 1.29. Accordingly, this part of i
the allegation is of concern. With regard to the suspended ceiling and l J-46 I I
lighting supports, the acceptability of the installation will depend on the approval by the TRT of the analysis to be provided by TUEC concerning the adequacy of the seismic Category II restraints in the control room. This technical issue, including the nonsafety-related conduit support system, will be resolved after the review of TUEC's seismic analysis substantiating the adequacy of the overall seismic support system installa-tion in the control room. The results of the TRT review of 1UEC's analysis will be reported in a supplement to this SSER. Based en the review of other seismic Category I areas of the plant, the TRT concludes that the acceptability of the installation will depend on TRT approval of TUEC's analysis of the adequacy of the seismic support installation for nonsafety related conduits in areas of the plant other than the control room. The TRT further concludes that the lack of analysis to substantiate the adequacy of the seismic design installations inspected may be an indica-tion of weakness in the QA/QC program concerning design control. This area is addressed under the.QA/QC Category 1, " Design Process."*
- 5. Action Required: TUEC shall perform the following actions prior to fuel load:
(a) Provide the TRT with analyses that substantiate (1) the adequacy of the overall seismic support system installation for all the items located above the ceiling in the control room, including nonsafety-related conduit, suspended ceiling, and lighting fixtures and (2) the adequacy of the seismic support system installation for nonsafety-related conduit in seismic Category I areas of'the plant other than the control room. This action shall be integrated as appropriate with other actions addressed under Civil and Structural Category 14,
" Seismic Design of Control Room Ceiling Elements."
(b) Evaluate the adequacy of the QA/QC program related to the deficiencies identified above to establish root causes and appropriate actions. These actions should be integrated with other actions addressed under the QA/QC Category 1, " Design Process." l l l i RThe TRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER. J-47
- 1. Allegation Category: Electrical and Instrumentation 5, Electrical Nonconformance Report (NCR) Activities
- 2. Allegation Number: AQE-1, AQE-2, AQE-3, AQE-4, AQE-5, AQE-25, AQE-33, AQE-34, AQE-35, AQE-37, AQE-38, AQE-40, AQE-41, AQE-42, AQE-45, AQE-47, AQE-48, AE-24, and parts of AE-22, AE-27, AQE-12, AQE-36 and AE-50.
- 3. Characterization: It is alleged that the validity of the generation and disposition of electrical nonconformance reports (NCRs) was suspect.
- 4. Assessment of Safety Significance: The implied safety significance of these ' allegations is that the quality of the electrical installation could be indeterminate.
These allegations pertain to various concerns involving the NCR program, and include: Prevalent "use-as-is" dispositions of NCRs (AQE-33, AQE-47, AQE-34, AQE-35, Parts of AE-27 and AQE-36). Inaccurate evaluation in the generation of NCRs to indicate workman-ship not compromised (AQE-48), The closing out of NCRs by unqualified inspectors (either inte'n-tionally or under coercion) (AQE-4). Pressure not to generate NCRs (AQE-42). The traceability of "Q" items (non-Q fuse blocks were installed where Q blocks were required) (AQE-35). Restraint cable (mechanical) crimp gauge calibration (AQE-41). Failure to follow procedures, specifications, and drawings (AQE-25, AQE-40, part of AQE-12). Splicing of safety-related electrical cables in violation of regula-tory requirements (part of AE-50). Questionable dispositions for NCRs involving inadequate thread engagement on a conduit fitting and damaged cable (AQE-45). Electrical cable tray fell, damaging cables entering the control room (AE-24). No documentation available for butt splices in panels (part of AE-22). Conduit replaced in Fuel Handling Building was dispositioned as repaired rather than replaced (AQE-3). In addition to these general concerns, several allegations contained specific information about questionable NCR dispositions, which includes: J-49
- Improper documentation was kept for the removal and pulling of damaged cables (AQE-1, AQE-2, AQE-5).
- Disposition of NCR on terminal block rework was questionable (AQE-37).
- Excessively bent terminal lugs in motor control centers (part of AQE-36).
- Unauthorized solenoid repair (AQE-38).
- Loose elbow termination conduit fittings found on the Unit 1 diesel generators (part of AE-27).
The NRC Special Review Team (SRT) also had concerns with respect to the Texas Utilities Electric Company (TVEC) management response to the so-called "T-shirt" incident because of its potential effect on the morale of QC electrical inspectors, which in turn could have affected their workmanship. (For detailed discussion of the "T-shirt" incident, see QA/QC Category 6, "QC Inspection," AQ-46.*) The Final Safety Analysis Report (FSAR), Section 17.1, " Quality Assurance During Design and Construction," commits TUEC to a quality assurance (QA) program, as required by 10 CFR 50, Appendix B. FSAR Section 17.1.10, "In-spection," outlines the inspection plans which will ensure that construction tasks conform to procedures, drawings, specifications, codes, standards, and other documentation. These plans are augmented by TUGC0 procedure CP-QP-16.0, which established the methods for generating and dispositioning reported items of nonconformance. The NRC Technical Review Team (TRT) reviewed pertinent TUEC documentation to determine that the procedures and instructions for generating and dispositioning reported items of noncon-formance were adequate as related to the concerns raised by the allegations. The TRT reviewed a random sample of 75 electrical NCRs and conducted numerous interviews with QA/QC and engineering personnel. (See also Electrical and Instrumentation Category 6, " Electrical QC Inspector Training / Qualifications.") The TRT reviewed 25 of the 75 electrical NCRs to determine if the QC inspectors who " closed out" the NCRs were qualified to do so. The TRT found that in all 25 cases the QC electrical inspectors were qualified and their certification files were current (AQE-4). Equipment installation matters raised by these allegations are addressed under: ,
- Electrical and Instrumentation Category 1, " Electrical Cable Termi-nations," for parts of AE-22, AQE-12 and AQE-36.
- Electrical and Instrumentation Category 2, " Electrical Cable Tray and Conduit Installation," for the alleged loose conduit fittings for part of AQ-27.
*The TRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER.
J-50
l Electrical and Instrumentation Category 7, " Electrical Cable Installa-tion," for the alleged splicing of safety-related cables in raceways and cable damage where trays contained trash and hazardous debris, for part of AE-50 and AQE-5. The TRT interviewed a TUEC electrical engineer and a lead quality engi-neer (QE) about the "use-as-is" disposition of electrical NCRs (AQE-33, AQE-47, AQE-34, AQE-35 and parts of AE-27, and AQE-36). The TRT deter-mined that for an NCR to receive a "use-as-is" disposition, an independent verification inspection by an electrical engineer had to be made for each reported item of nonconformance. Based on that inspection, and on an evaluation with regard to procedures, specifications, drawings (including applicable codes and standards), and other related documentation, a "use-as-is" disposition could be applied. Final approval of such a disposition required two QE signatures. The TRT also reviewed the 75 NCRs to determine if there were any with the disposition "use-as-is" with the explanation "not addressed in ES-100," as alleged. The TUEC engineer indicated that should an NCR be received with this type of disposition, it would be " kicked back" and would require more justification. The TRT determined that if the nonconformance indeed was not addressed in ES-100, then a document, such as a procedure or other specification, that did address this nonconformance item would be required to be refer-enced in the NCR. Of the 75 NCRs examined, the TRT could identify no "use-as-is" dispositions which deviated from applicable design require-ments, except for those identified in Electrical and Instrumentation Category 1, " Electrical Cable Terminations," and Electrical and Instru-mentation Category 2, " Electrical Cable Tray and Conauit Installation," regarding NCRs identifyin'g bent terminal lugs in motor control centers (part of AQE-36) and reporting two loose conduit elbow fittings (part of AE-27), respectively. These TRT findings were discussed with the allegers, one of whom disagreed with the TRT findings as related to AQE-34 and AQE-35 and provided additional information. The TRT is currently evaluating this new information and will report its findings in a supplement to this SSER. The TRT also interviewed a TUEC electrical engineer about NCR dispost-tions with respect to " replace versus repair" (AQE-3) and " compromised workmanship" (AQE-48). The TRT determined that replacing a reported item instead of repairing it as originally dispositioned would require a revision to the original NCR. The disposition of the NCR for replacement would be based on an engineering evaluation. The TRT determined that on a case-by case basis where workmanship might have been compromised, the inspecting engineer would apply engineering judgment to determine that the quality of workmanship did not degrade the installation below an acceptable level. From the 75 NCRs examined the TRT could not find any evidence of unacceptable installation. (See also Electrical and Instru-mentation Category 8, " Electrical Procedures," regarding correction of installation deficiencies for lighting terminations.) The TRT searched the records for the number of NCRs and inspection reports written and for the amount of cable pulled for a 57-day period prior to and a 57-day period following the so-called "T-shirt" incident. This search was conducted to determine if the incident had any effect on i J-51 !
the workmanship of the electrical QC inspectors. The TRT could find no evidence that inspectors were affected by the incident as a result of management reaction to it. The TRT interviewed the quality control (QC) supervisor of the calibra-tion lab and reviewed pertinent prucedures that were followed to ensure that construction tools which required periodic calibration were maintained (AQE-41). The TRT found that lab controls, procedures, and tool traceability, if properly implemented, would ensure that tool cali-bration was maintained. Adequate procedures also existed to ensure that corrective actions would be taken if a tool did not meet calibration t specifications and tolerances. The TRT reviewed NCR documentation on tool calibration and found it to have been dispositioned in accordance with procedures that ensured the integrity of the construction (See also QA/QC Category 6, "QC Inspection," for the disposition of the specific concern raised by AEQ-41.) The TRT interviewed QC and purchasing personnel and an electrical general foreman for construction and reviewed pertinent documentation to determine the adequacy of traceability of safety-related (noted as "Q") items (AQE-35). The TRT determined that procedures and controls, if properly followed, were adequate to ensure the traceability of "Q" items and that they would preclude the possibility of substituting "non-Q" for "Q" items. The TRT reviewed a large number of installation documents and found all the required traceability documentation. In regard to AQE-42, the TRT's interviews with QC personnel could not substantiate the allegation that an individual was pressured not to issue NCRs. (See also QA/QC Category 6, "QC Inspection," AQ-35.) To address the specific technical concerns raised in the above allega-tions, the TRT examined the NCR log books, interviewed allegers, and selected a random sample of NCRs pertaining to specific items of concern. The TRT determined that:
- The allegation (AQE-36) of excessive bending of AMP Product Corporation compression lugs in ITT Gould-Brown Boveri switchgear was substantiated. This issue is addressed in Electrical and Instrumentation Category 1, " Electrical Cable Terminations."
- The allegations of improper documentation of cable removal (AQE-1 and AQE-2); repair rather than replacement of flex conduit (AQE-3); damaged cable as a result of a fallen cable tray (AQ-24);
failure to follow procedures and specifications (AQE-25 and AQE-40); damaged cable due to inadequate thread engagement on a conduit (AQE-45); and rework of terminal blocks (AQE-37) could not be substantiated, since in its review of a random sample of 75 NCRs on these issues the TRT could not identify any inconsistencies or These findings deficiencies that would raise a safety question. were discussed with some of the individuals responsible for raising these concerns, one of whom disagreed with the TRT determination concerning AQE-37 and provided add,itional information. The TRT is currently evaluating this new information and will report the results in a supplement to this SSER. J-52
The alleger clarified AQE-38 during an interview with the TRT, indicating that the concern related to the repair of an off-the-shelf solenoid (glueing the connecting terminal to the solenoid coil and resoldering the coil lead to the terminal). The alleger believed that the scienoid was used in a safety-related system, but could not remember which system. Moreover, the alleger indicated that there was no written record of the repaired solenoid. The TRT could not substantiate the concern raised by this allegation. The allegation concerning the ~ nose elbow termination conduit fitt-ings in the diesel generator rov: s for Unit 1 (Part of AE-27) has merit. The TRT examined the NCR log book and found the specific NCRs for this item. The TRT also inspected the diesel generator rooms of Unit 1 and found two loose elbow conduit fittings. This issue is addressed in Electrical and Instrumentation Category 2,
" Electrical Cable Tray and Conduit Installation."
- 5. Conclusions and Staff Positions: Based on the reviews of the' pertinent documentation, examination of NCRs, and the information obtained from the interviews, the TRT concludes that adequate procedures, controls, and process checks exist for the generation and disposition of reported items of nonconformance as related to the concerns raised by the above allegations. The TRT also concludes that of the allegations identified at the outset of this section, only a few specific instances were found which raised questions concerning the adequacy of safety-related items.
These are discussed above and are dis' cussed further in other sections of the report. The results of this evaluation will be further assessed as part of the overall programmatic review of all NCRs, addressed under QA/QC Category 5, "Nonconformance Reports," and under QA/QC Category 6, "QC Inspection." Therefore, the final acceptability of this evaluation will be predicated on the satisfactory result of the overall programmatic review on these subjects. Any adjustments to these conclusions will be reported in a supplement to this SSER. The results of the TRT review of new informa-tion concerning allegations AQE-34, AQE-35 and AQE-37 will also be reported in a supplement to this SSER.
- 6. Action Required: None.
! J-53 l l t
- 1. Allegation Category: Electrical and Instrumentation 6, Electrical Quality Control (QC) Inspector Training / Qualifications
- 2. Allegation Number: AQE-8, Parts of AQE-4 and AQE-12.
- 3. Characterization: It is alleged that some electrical QC inspectors were inadequately qualified, that they received help in passing certification tests, and that their previous experience was inadequate to fulfill the job requirements.
- 4. Assessment of Safety Significance: The implied safety significance of these allegations is that the lack of training or qualification of electrical QC inspectors could result in inadequate inspections of safety-related components.
The allegations question whether the positions of American National Standards Institute (ANSI) Standard N45.2.6-1978, " Qualifications of Inspection, Examination, and Test Personnel for the Construction Phase of Nuclear Power Plants," as augmented in the Final Safety Analysis Report (FSAR) Section 17.1.2, " Quality Assurance Program," were con-sidered by Texas Utility Electric Company (TUEC) in the development of the quality assurance (QA) program at the Comanche Peak Steam Electric Station (CPSES). Regulatory Guide (RG) 1.58, Revision 1, " Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel," endorses the positions of ANSI N45.2.6-1978. RG 1.58, Revision 1, and ANSI N45.2.6-1978 set forth positions stating the education and experience requirements for the various capability levels of inspectors (I, II, and III). Both documents, however, state that these requirements are not absolute when other factors may provide reasonable assurance that a person can competently perform a particular task. They require that all records or qualifications shall be maintained by TUEC in an individual's personnel file. In assessing these allegations, the NRC Technical Review Team (TRT) examined Texas Utilities Generating Company (TUGCO) procedures, QC inspector training and certification files, testing program requirements, on-the-job training (0JT) requirements, and recertification program requirements. The TRT also conducted interviews with the training coordinator, two Level I QC electrical technicians, four Level II QC electrical inspectors, one Level III quality engineer (QE), one Level II lead QC electrical inspector, one lead QE, and the QE Supervisor. Procedures. The TRT found that TUGC0 Procedure CP-QP-2.1, " Training of Inspection Personnel," commencing with Revision 8 (July 1981), contained 3 l education and experience requirements consistent with RG 1.58, Revision 1, l and ANSI N45.2.6-1978. Revision 7 (June 1981) of the above procedure, ! Section 3.1.d, " Technical Training" contained the statement: Minimum training, education, and experience requirements will be defined in technical training outlines prepared for specific inspection activities (civil, electrical, etc.). J-55
e After a discussion with the training coordinator and an examination of the technical training outlines, the TRT discovered that the education and experience requirements were never defined and that only the training requirements had been defined. After examining other related procedures, the TRT found the following deficiencies. Trainina and Certification Files. The TRT examined in detail six elec-trical QC inspectors' training and certification files (two Level I and four Level II). The examination revealed the following two instances where TUGC0 Procedure CP-QP-2.1, Revisions 8 through 15, RG 1.58, Revi-sion 1, and ANSI N45.2.6-1978 requirements for qualification were not being met: l I (a) There was no documentation of a high school diploma or General Equivalency Diploma (GED) for one of the inspectors selected. The file on this inspector contained only a telephone conference note that a call had been made in 1982 requesting information from a high l school. l (b) There was no documentation to waive the remaining 2 months of the I required 1 year of experience for a Level I technician before the l individual became a Level II inspector after successfully passing the required examinations. The TRT also found one case where a Level I QC technician had no't passed the required color vision examination, which was to be administered by an independent professional eye specialist. A makeup test using colored pen-cils was administered by a QC supervisor, was passed, and then a waiver was given. ~A TUGC0 procedure allowed for a waiver on a case-by-case basis. In addition to the above, the TRT staff also found two cases where the ex-perience requirements to become a Level I technician were met only marginally. In one case, no documentation was found in the training and certification files substantiating that the person met the experience requirements or providing the basis for determining that the person could, with reasonable assurance, competently perform the particular task without having the required related experience. Testina Program Requirements. The TRT examined the testing, ratesting and scoring methods applicable to Level II qualification and found some guide-line inconsistencies and procedural deficiencies. Specifically,.they included: (a) No time limit or additional training requirements between a failed test and a retest. In practice, the time varied from a few days to months. (b) No controls to assure that the same test would not be given if the taker previously failed it. (c) No consistency in scoring. Two different scoring techniques were used to-average the results when two tests were taken. Combined test. scores could vary slightly, depending on which technique was used. These slight variations could make the difference between passing or failing the tests - - a condition resulting solely from J-56
._ .. -- - - -.~.- - . - - ... ~ - . ~-
f ;_. r ~ I the scoring technique used. Seven out of 25 tests used one test 1 scoring technique instead of the other. : i
- 1 (d) No guidelines or procedures to control the disqualification of ques-tions from the test. In one. instance a question was disqualified after the test was administered, thus allowing two people to pass the exam that they would have otherwise failed.
(e) No program for establishing new tests (except when procedures changed). i The same tests had been utilized for the last 2 years. . i On-The-Job (0JT) Training Requirements. The TRT examined the OJT training for QC electrical inspectors and found sufficient documentation in the j training and certification files that adequate OJT was being obtained. j Numerous cases were found where a portion (10%-20%) of the required OJT j was being waived only after applicants successfully passed the Level II ] examinations. i Recertification Program Requirements. The TRT examined the recertification 1 ' program and found that there was no required documentation to assure that , t recertification requirements were being met. The present system only requires a simple "yes" or "no" answer from an inspector's lead QC inspec- ' l tor that the individual had been active in the area in the last 6 to 12 - months and was knowledgeable about current procedure requirements. The 3 lead QC inspectors did not maintain any written record of a subordinate inspector's activity. -- Interviews. The TRT interviewed 11 people, including the training coordi-nator and Level-I QC technicians on up to the QE supervisor. The con- - sensus of those interviewed was that the training program was adequate and ! had improved over the last couple of years. Some thought additional OJT l would have been more beneficial in lieu of " book time." l Based on reviews of the QC inspector training and qualification aspects of
- the electrical QA program, the TRT determined that current procedures in 4
effect beginning with Revision 8 of the CP-QP-2.1' meet the requirements of ! ANSI N45.2.6-1978, as augmented in the FSAR and endorsed by RG 1.58, Revi- ! sion 1. Prior to Revision 8, TUGCO procedures did not define the education l and experience recommended in the above regulatory documents. TUGC0 was not l committed to these requirements until April 30, 1981. The TRT review of the j training and certification files determined that some supportive documenta- { tion, as required by procedures and regulatory positions, was lacking. The TRT determined that the testing program lacks guidelines and procedural requirements covering, but'not limited to, such items as test question dis-qualifications, scoring, retests, and the prolonged use of the same tests. l- The TRT also determined that the inspector recertification program lacks i programmatic controls to assure that the recertification requirements in ~ the different electrical quality instructions are being met. l S. Conclusions and Staff Positions: Based on its review of the pertinent _;
- documentation and its interviews, the TRT concludes that there is evidence i to indicate that the electrical QC inspector qualification program lacked j programmatic controls, which may be indicative that the required level of i
J-57 i
_ _ _. _.._-- . __ _ _ _ _ _ . _ - . . _ - . - _ ~- ._ .- i Spe-qualification was not obtained for some electrical QC inspectors. cifically, the lack of programmatic controls to assure that suitable proficiency is achieved and maintained (as required by 10 CFR 50, Appen-i dix B) was found in:
- The supportive documentation of qualifications, as required by pro-cedures and regulatory requirements in the training and certification.
- The testing program for Level II qualification.
- The recertification program requirements in electrical quality i instructions.
1 The TRT concludes that the lack of these programmatic controls in the i, electrical QC inspector qualification program is of concern. l Since the training and certification program is the same for all disci- { j plines (except ASME), the TRT concludes that the deficiencies in procedural requirements and guidelines in the testing program and the lack of documen-tation in isolated cases have generic implications to the other construc-tion disciplines. The implications of the TRT's findings concerning
; electrical QC inspector training and oualification will be further assessed as part of the overall programmatic review of QC inspector training and i
qualification, which is addressed under QA/QC Category 4, " Training and Qualification."* l 6. Action Required: TUEC shall accomplish the following prior to fuel load: l (1) Evaluate the testing program for QC electrical inspector qualifica-i tions and develop a testing program which optimizes administrative i guidelines, procedural requirements, and test flexibility (e.g. , j computer generated tests) to assure that suitable proficiency is achieved and maintained. These guidelines and/or procedures shall include such items 'as scoring, ratests, and question disqualification. 4 (2) Justify the allowance in the procedure for administering separate (waiver) vision tests in lieu of examinations administered by'an l independent professional eye specialist.
~
(3) Review all electrical QC inspector training, qualification, certifi-cation and recertification files against the project requirements as documented in the FSAR, and provide the information in such'a form that each requirement is clearly shown to have been met by each inspector. If an inspector is found to not meet the training, i qualification, certification, or recertification requirements, TUEC shall then review the records to determine the adequacy of inspec-i tions made by unqualified individuals and provide a statement on ' j the impact of the deficiencies noted on the safety of the project. 1 (4) Integrate these actions, as appropriate, with other actions addressed 1 I under QA/QC Category 4, " Training and Qualifications." j i
*The TRT evaluation of QA/QC allegations is in progress and will be published 4 in a subsequent supplement to this SSER.
J-58 l 1
- 1. Allegation Category: Electrical and Instrumentation 7, Electrical Cable Installation
- 2. Allegation Number: AE-19, AE-28, AE-30, AE-50,_ Parts of AQE-5 and AE-29 and Special Review Team SRT-10
- 3. Characterization: It is alleged that:
- Cable trays were overloaded (AE-19).
Cables were not " trained" in a workmanlike manner in the cable spread-
- ing room and in junction boxes 1058 and 1059 (AE-28).
Higher siderails were added to cable trays due to tray overfill con-
- ditions (Part of AE-29).
Cable density / compaction problems may exist due to tray overfill
- conditions (AE-30).
Cables were spliced in cable trays in the cable spreading room in
- violation of regulatory requirements (AE-50).
A nonconformance report pertaining to trash in cable trays, damaged cable, and improperly trained cable was improperly closed (Part of AQE-5). The Special Review Team Report on July 13, 1984, identified the issue of ' overloaded cable trays due to the installation of "thermolag" material (SRT-10).
- 4. Assessment of Safety Significance: The implied safety significance of these allegations is that improperly trained cables, improper cable splices and overloaded cable trays could place the quality of the installation in question. -
Cable Splices in Raceways. Allegation AE-50 involved the alleged splicing of safety-related ments. cables in raceways in violation of regulatory require-The NRC Technical Review Team (TRT) reviewed NRC Region IV (RIV) inspection report 83-03 (November 8, 1982) and found that the RIV investigation of the two cables specifically identified by the alleger adequately addressed this allegation. The RIV investi Gation determined that one cable no longer performs a safety related function, and the other cable had become a " spare" and was removed from the raceway. The TRT determined that similar-appearing items in the same area were not splices, but were, in fact, acceptable methods of repairing minor cable jacket damage. The TRT concurs with the RIV determination but notes that regulatory requirements discourage the use of splices in raceways, as stated in position 9 of Regulatory Guide (RG) 1.75, " Physical Independence of Electric Systems." If splices are made, the resulting design should be justified by analysis. Category 8, "As Built."" This area is further addressed under QA/QC' The TRT examined the cable spreading room, identified two cables installed in raceway, which to'the untrained eye could appear to have been spliced, and inspected them in their as-installed condition. The TRT~also reviewed the applicable installation / inspection records. This inspection and QThe TRT evaluation of QA/QC allegations is in' progress and will be published in a subsequent supplement to this SSER. J-59
l
\
review revealed that there were cable jacket repairs and that they were properly identified, repaired, and documented in accordance with applicable , procedures. Poor Workmanship. A11egation'AE-28 and part of AQE-5 involved instances of improper cable " training" (or dressing), poor workmanship in cable installation, and cables installed in raceways containing trash and hazard-ous debris. The issues of improper " training" of cables and poor workman-The TRT ship in junction boxes 1058 and 1059 were inspected by the TRT. findings agree with the previous NRC RIV determination that these cables, which are n'onsafety-related, were properly trained and that they exhibited an acceptable degree of workmanship. These findings were discussed with the alleger who indicated that the junction box numbers may not have been correct and provided additional information concerning the location of the boxes in the plant. The TRT is currently evaluating this new The information alleger did and will report the results in a supplement to this SSER. not identify which trays contained trash and hazardous debris at the time of cable installation, so the TRT randomly inspected approximately 2,000 feet of cable trays containing safety-related cables and found no l instances of improper training, trash, hazardous debris, or poor l workmanship. Tray Overfill. Allegations AE-19 and AE-30 involved various concerns related to cable trays possibly being overfilled. The alleger The TRT specifically inspection identified tray T130CC007 in the cable spreading room. of this tray revealed the following: (a) Siderails were installed on this tray, adding approximately 2 inches to its height. When inspected, no cables extended above the level of the siderails. (b) Per nonconformance report (NCR) E-82-1073R1, eight spare cables were removed from this tray in January 1983, in conjunction with the removal of 42 spare cables from tray T130ECC82 because the tray was identified overloaded. (c) Calculatiun of the actual weight of cables currently installed in this tray indicated loading of approximately 22 pounds per square foot, compared with the maximum allowable value of 35 pounds per square foot, as specified in seismic supporting requirements. (d) Calculation of the square area fill of cables currently installed in this tray indicates an actual fill of 28%, compared with the maximum recommended value of 40%, as stated in IEEE Standard 422, " Guide for
' the Design and Installation of Cable Systems in Power Generating Stations." The TRT selected nine additional sections of tray con-taining large quantities of cables. These quantities ranged from 57 to 300 cables per tray section. The square area fill and weight per square foot values for these trays were reviewed for conformance with the stated maximum values. The results of this review were as follows:
(1) All nine trays were loaded at less than 28 pounds per square foot. J-60
(2) Seven of the trays had square area fill less than 40%. (3) The two remaining trays had square area fills of 41% and 42%; however, Section 8.3.3.1 of the Final Safety Analysis Report (FSAR) justifias exceeding the 40% value if cables do not extend above the siderails of the tray, and do not violate seismic supporting requirements. The NRC staff considers this justification acceptable. This review revealed that all trays sampled comply with seismic supporting requirements and, because no cables extended above the tray siderails, that no deficiencies existed within the sample selected. Added Loads on Trays. Part of allegation AE-29 and concern SRT-10 involved the addition of higher siderails and "thermolag" material to existing cable trays, conditions which could cause trays to become physically overloaded. Regarding the higher siderails, the TRT discovered that siderails were fabricated using 6-inch high by 16 gauge galvanized sheet metal. As such, the addition of this material would increase tray loading by approximately 2\ pounds per foot. Using the above sample of cable trays, which the TRT considers representative of some of the most highly loaded trays at Comanche Peak Steam Electric Station (CPSES), Unit 1, this added height would bring the most highly loaded tray to approximately 30.5 pounds per square foot, compared with the maximum allowable value of 35 pounds per square foot. Regarding the "thermolag" material, the TRT reviewed procedure CP-EI-4.0-49, Revision 1, " Evaluation of Thermolag (TSI) Fire Barrier Material on Class 1E Electrical Raceways." From this review the TRT determined that the procedure was adequate if properly followed to assure that, should over-loading occur due to the addition of thermolag material, these instances would be identified, evaluated, and if necessary, corrected prior to the installation of the thermolag. The TRT then selected two raceways (one cable tray and one conduit) with thermolag installed and reviewed the evaluations performed in accordance with the above procedure. The TRT found that the requirements of the procedure had been met, and therefore, determined that the addition of tray siderails and thermolag material poses no hazard to the structural integrity of the raceway system.
- 5. Conclusions and Staff Positions: Based on the inspection of the cable installations for cable splices in cable trays, workmanship, cable tray fill, added load on cable trays by thermolag material, and review of pertinent criteria, procedures, RIV inspection reports, installation /
inspection reports, and NCRs, the TRT concludes that the various aspects j of the cable installation on raceway fill reviewed and inspected meet established installation requirements. Therefore, the TRT concludes that these allegations could not be substantiated. The results of the TRT review of new information concerning allegation AE-28 will be reported in a supplement to this SSER.
- 6. Action Required < None.
l l l J-61
- 1. Allegation Category:
Procedures Electrical and Instrumentation 8, Electrical
- 2. Allegation Number:
AQE-23, AQE-32, AQE-39, AQE-44, AQE-46, AQE-52 and Parts of AQE-6, AE-18 and AE-20.
- 3. Characterization. It is alleged that:
Requirements were deleted in the procedural revision for post-construction inspection of electrical equipment and raceways and electrical QC inspectors were directed by a supervisor not to follow procedures (AQE-23 and part of AQE-6). The number of required inspections was reduced in the inspection procedure for reverification of seismic electrical equipment mounting details (AQE-32). Revisions to the procedure for post-construction inspection of electrical equipment and raceways were made to accommodate numerous problems with loose terminations found in the lighting system termi-nal boxes during past inspections (AQE-39 and AQE-46). Revision 15 to the procedure for post-construction inspection of electrical equipment and raceways omitted requirements for inspections of large pieces of equipment, such as 6.9-kilovolt (kV) motors (AQE-52). A cable separation problem identified in the Fuel Handling Building was dispositioned "use-as-is," contrary to procedure (AQE-44). Insulated butt splices were being used inside panels in violation of the in process inspection procedure for cable terminations (Part of AE-18). Separation criteria between redundant cable trays and conduits in the cable spreading room were not consistent with the requirements of the in process inspection procedures for verifying electrical separation (Part of AE-20).
- 4. Assessment of Safety Significance: The implied safety significance of these allegations is that the quality of the electrical installation may be in question because requirements were deleted from procedures, required inspections trary were reduced in frequency, and installation was being done con-to procedures.
The NRC Technical Review Team (TRT) examined nine in process inspection procedures used during plant construction, one post-construction inspec- l tion and walkdown procedure, and four turnover inspection procedures for final acceptance of station systems, structures, and equipment by TUEC startup and operations. The TRT reviewed in place procedures, historical procedure files, inspection reports (irs). IR deficiency logs, post-construction deficiency lists, electrical equipment punch lists, electrical separation deficiency reports, test release / return to contractor custody /startup release to operations forms, construction J-63
operation travelers, startup work authorizations, and systems /areaThe TRT also in testing, drawing, walkdown results/ review forms. The TRT examined the above QC management personnel and allegers. documents for programmatic we may have negated quality assurance / quality control (QA/QC) inspection activities during construction of the plant. Procedures for Post-Construction Inspection of The Electrical Equipment TRT review and of proce-Raceways (AQE-23, AQE-52 and Part of AQE-6). dure QI-QP-11.3-40, " Post-Construction Inspection of Electrical Equipment and Raceways," revealed that most deficiencies identified by QA/QC personnel during post-construction and walkdown inspections of electrical equipment and raceways was based on this procedure, which provides ade-quate guidance for electrical equipment and raceway inspections. walkdown procedure had under-The TRT found that this post-constructionBefore Revision 15, QC inspectors w gone 18 revisions. procedure extensively to reinspect in process inspection activities (e.g., Sections 3.1.1 through 3.1.4 of Revision 14, requiring verification of cable, cable tray, conduit, and equipment installation, which were re-written under Sections 3.1.1 and 3.1.2 of Revision Revision15, 15,entitled, Section" 3.1.2,Raceway Inspection" and " Equipment Inspection"). 2 J There-covers requirements for inspection of large Class 1E equipment. fore, 6.9-kV motors, considered to be large equipment, However,would sincehave they been are not covered by the procedure, if classified IE. Class IE, they could be excluded from inspections. Some of the revisions of this procedure came as a result of the many test deficiency change requests (TDCRs) based on TUGC0 procedure CP-SAP-3,
" Custody Transfer of Station Components." These deficiencies evolved from the startup performance testing of components and systems that B&R and other contractors had turned over to TUGCO. Other revisions were made include the experience gained during the reinspection of the in process inspection activities.
After a review of QI-QP-11.3-40 and CP-SAP-3, as well as other pertinent electrical in process inspection and startup administrative procedures, the TRT did not find any omissions in requirements for inspection of electrical equipment and raceways (AQE-23 and part of AQE-6). The Procedures for Lichtina Termination and Wirina (AQE-39 and AQE-46 TRT found that safety-related lighting terminations and wiring were required to be inspected under TUGC0 in process procedures QI-QP-11.3-23,
" Class 1E Conduit Raceway Inspections," QI-QP-11.3-26, " Electrical Cable Installation Inspections," QI-QP-11.3-28, " Class IE Cable Terminations,"
and QI-QP-11.3-40, " Post-Construction Inspection of Electrical Equipment and Raceways." The TRT found that the inspections of emergency lighting and associated terminations were being performed under Revision 15 or earlier revisions of procedure QI-QP-11.3-40, even though the procedure Revision 16 ofwas not specifica this pro-addressing the emergency lighting inspections. cedure was made specifically to address raceway lighting inspections (Section 3.3.1).
)
J-64
The TRT found that the loose terminations within the lighting termination boxes occurred as a result of an installation deficiency by craft personnel involving the Thomas and Betts RP-12 crimp-type insulated connectors. A document change notice (DCN) was issued changing the engineering instruc-tion used by craft personnel (EE-8) to improve installation of lighting terminations; thereafter the number of deficiency reports in lighting termination boxes was greatly reduced. (See also QA/QC Category 8, "As Built," for conclusions regarding craft personnel training.) The TRT found that the revisions to procedure QI-QP-11.3-40 regarding emergency lighting inspections were justified to eliminate unnecessary inspection requirements. Other Electrical Procedures (AQE-32, AQE-44 and Parts of AE-18 and AE-20). After a review of procedure QI-QP-ll.14-12, " Reverification of Seismic Electrical Equipment Mounting Details," the TRT could find no requirements in Revision 0 through 4 that established a fixed frequency for reverifica-tion of inspections concerning bolt tightening of seismic electrical equipment mountings. However, the procedure provided for reverification of inspections on a " case-by-case" basis (AQE-32). The TRT also reviewed the following in process inspection procedures with respect to electrical equipment separation and the use of butt splices in panels (parts of AE-20 and AE-18): (a) Procedure QI-QP-11.3-29, " Electrical Separation," (b) Procedure QI-QP-11.3-29.1, " Verify Electrical Separation," (c) Procedure QI-QP-11.3-28, " Class 1E Cable Terminations." The TRT determined that in process inspection procedures QI-QP-11.3-29 and QI-QP-11.3-29.1, and post-construction procedure QI-QP-11.3-40, were used to identify deficiencies in the Fuel Handling Building and that these procedures allow the "use as-is" disposition of nonconformance reports (NCRs). The subject of "use as-is" disposition of NCRs (AQE-44) is dis-cussed in Electrical and Instrumentation Category 5, " Electrical Noncon-formance Report (NCR) Activities." The separation of electrical equipment and installation of terminations in accordance with procedures, drawings, and specifications are discussed in Electrical and Instrumentation Category 1, " Electrical Cable Termina-tions," for part of AE-18 and Electrical and Instrumentation Category 3,
" Electrical Equipment Separation," for part of AE-20.
In a TRT review of otner electrical procedures, the TRT found no omissions in requirements for inspection of electrical equipment.
- 5. Conclusions and Staff Positions: Based on its review of procedures for in process inspections, post construction, and turnover inspections, the TRT concludes that no significant concerns exist with electrical proce-dures. However, equipment installation problems as related to non-conformance with procedures are being addressed in the hardware-related E&I categories. The TRT, therefore, concludes that these electrical procedure related allegations could not be substantiated.
J-65
l The results of this evaluation will be further assessed as part of the overall programmatic review concerning the post-construction verification Therefore, the program addressed under QA/QC, Category 8, "As Built."* final acceptability of this evaluation will be predicated on the satis-Any factory results of the overall programmatic review on this subject. adjustments to these conclusions will be reported in a supplement to this SSER.
- 6. Action Required: None.
*The TRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER.
J-66 i
1 l l J 1
- 1. Allegation Category: Electrical and Instrumentation 9, Electrical Inspection Reports, Inspection Item Removal Notices and In-Process Inspections.
- 2. Allegation Number: AQE-7 and AQE-43
- 3. Characterization: It is alleged that the per procedure number of required 2
in process inspections was not being conducted and that inspection reports I (irs) were.being written without reinspections to close out inspection item removal notices (IRNs). 4. Assessment of Safety Significance: The implied safety significance of these allegations is that a reduction of in process inspections and an omission of reinspections could compromise the quality of the installa-tion of safety-related components. In-Process Inspections (AQE-7). The NRC Technical Review Team (TRT) examined current and past inspection procedures in the electrical area to determine the number of in process inspections required. The TRT found that Texas Utilities Generating Company (TUGCO) procedure QI-QP-11.3-28, " Class IE Cable Terminations," was the only electrical inspection procedure which defined a specific number of required in process inspections. Through Revision 4 (dated July 16, 1980), the procedure required a minimum of 10 in process inspections per shift;
- revision 5 of the procedure (August 7, 1980) changed the quantity required to "a weekly" in process inspection.
The TRT interviewed quality control (QC) personnel to learn the basis for i the substantial revision to the procedure. However, the individuals responsible for this revision were no longer employed at Comanche Peak Steam Electric Station (CPSES) and could not be contacted. Current QC personnel could only speculate that an increase in level of confidence was the basis for the reduction in inspections. The TRT interviewed the project engineering manager to determine the the amount of Class 1E cable termination activity at the time the procedure was revised. From the discussion, the TRT determined that there was less cable termination activity in early 1980 (before the procedure was revised) than in late peak. to mid-1981, when cable termination activity was approaching its 1980 Comparing the number of NCRs for cable termination activity for 2 years before revision 5 with the results of the quality assurance (QA) trend reports for 1980 (third and fourth quarters) and 1981 (first i and second quarters), the TRT determined that the number of NCRs for cable termination activity remained the same during this period, despite the much smaller number of in process inspections. This may be indicative { that the fewer inspections under revision 5 were much more thorough than i those before revision 5. However, the TRT could not substantiate the l improvement of the quality of the installation in view of the problems found with the electrical terminations discussed in Electrical and l Instrumentation Category 1, " Electrical Cable Terminations." Inspection Reports and Inspection Item Removal Notices (ACE-43). The TRT examined TUGC0 procedure CP-QP-18.0, " Inspected Item Femoval Notice Form," for its adequacy to control the inspection process. The TRT determined that this procedure was adequate to assure that reinspections J-67
were performed, when required, to verify thac the item subject to the IRN was still in conformance with the requirements. The TRT also interviewed two paper flow group (PFG) coordinators, a PFG IR clerk, and a lead QC electrical inspector, and examined 20 irs and IRNs. The TRT determined that because of the checking and paper pro-cessing involved with irs and IRNs, a PFG coordinator would not be able to recognize that a signed-off inspection report Afterhad been completed discussing this issue without reinspection actually occurring. with QC inspectors, the TRT determined that an inspection could be made without an inspection report in hand and after that inspection a report could be completed away from the inspection site, from which theThe TRT inference could be made that an inspection had not been made. found that there are no requirements in the procedures that inspection reports be in-hand before reinspections are conducted; hence, it can be construed that inspections may have been performed also without all required documentation in-hand. The TRT contacted the alleger, who Further, provided no additional information about the allegation. the alleger acknowledged when making the allegation and again during discussions with the TRT that this allegation was based on hearsay information. Based on its review of the pertinent
- 5. Conclusions and Staff Positions:
documents and interviews, the TRT concludes that the allegations about changing the frequency of in process inspections for cable terminations were unsubstantiated. However, cable termination problems that could be related to the concerns highlighted by these allegations are discussed in Electrical and Instrumentation Category 1, " Electrical Cable Terminations." The results of this evaluation will be further assessed as part of the overall programatic review of TUEC's deficiency identification program in process inspections addressed under QA/QC Category 5, "Nonconformance Reports."* Therefore, the final acceptability of this evaluation will be predicated on the results of the overall programmatic review of this sub-ject. Any modifications to these conclusions will be reported in a supplement to this SSER.
- 6. Action Required: The actions required in Electrical and' Instrumentation Category 1, " Electrical Cable Terminations," address the concerns with regard to reduction in cable termination inspections discussed above.
*The TRT evaluation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER.
J-68
- 1. Allegation Category: Test Program 1, Test Program Surfaced Numerous Deficiencies
- 2. Allegation Number: AT-1, 2, 3, 4, 5, 6, 8, 9, 10, 11, 13 and 17
- 3. Characterization: In support of a proposed contention (No. 26), the inter-venor, Citizens Association for Sound Energy (CASE), alleges that: (1) TUEC failed to conduct an adequate prefueling hot functional test (HFT) program, in that not all components or modifications were installed which require hot functional testing; (2) TUEC did not intend to check some components and systems until heatup to hot standby or during power escalation; (3) TUEC and the NRC Region IV staff failed to notice this condition and did not keep the ASLB-informed of the problems encountered; (4) numerous problems were identified during the conduct of the thermal expansion test, as evidenced by Test Deficiency Reports (TORS) 853 and 855; (5) the HFT was conducted without consideration of accident conditions; and, (6) TUEC and the NRC Region IV staff were willing to accept deficient test results.
For these reasons, CASE asserts that there was a lack of candor on the part of the NRC Region IV staff and TUEC and that the ASLB cannot rely on the NRC staff to monitor plant testing.
- 4. Assessment of Safety Significance: The implied significance of these allegations is that if the HFT program was improperly conducted, the adequacy of the plant to operate safely cannot be assured.
The NRC requires that a preoperational testing program on a nuclear power plant be conducted to demonstrate that plant structures, systems, and components meet their safety related design specifications, as stated in the utility's Final Safety Analysis Report (FSAR), before the plant goes into operation. The NRC Technical Review Team (TRT) conducted a review on 17 of 25 completed test data packages pertaining to HFT (which is a preoperational test) and interviewed cognizant TUEC personnel during the course of this review. The review included follow-up inspections on TDRs that were generated as a result of testing deficiencies found prior to and during HFT. The TRT also reviewed pertinent Startup Administrative Procedures, NRC Inspection Reports, the preoperational test index with schedule, and a Master System / Subsystem Index. The TRT reviewed this docu-mentation against the FSAR and the applicable NRC requirements and guidance (10 CFR 50 and Regulatory Guide 1.68) to evaluate TUEC's compliance. (1) The TRT confirmed that the HFT was conducted with some components and equipment not having been installed at the time of the test and with modifications remaining to be completed after the test. In order to determine whether TUEC had a technical basis for proceed-ing with the HFT when it was conducted, the TRT reviewed NRC Construc-tion Appraisal Inspection Report 50-445/83-18 (conducted January 24 , 1983 through February 4,1983), Inspection Report 50-445/83-23 (con-ducted May 23, 1983 through June 10, 1983) and Inspection Report 50-445/84-16 (conducted May 14, 1984 through June 20, 1984). This review was undertaken to determine if, prior to the start of the TRT's review effort, any NRC inspections around the time of the HFT had I i J-69 l
b identified missing components and equipment which had not been properly > documented in accordance with TUEC's established administrative con-trols. None were identified in those inspection reports. This pro-1
- vided a basis for the TRT to consider that those administrative controls had been properly-implemented through issuance of a TDR or Test Procedure Deviation (TPD). A TDR documents components and equip-l l
ment found to be deficient or defective at the time of the test and for which some action must be taken to correct the problem; a TPD docu- ' ments an approved change or deviation from the procedure as originally i- written. TDRs and TPDs become part of the completed test record which must be reviewed by the TUEC Joint Test Group (JTG) prior to its final acceptance of the test results. i Approximately 95 percent of the TORS issued relative to HFT documented piping and equipment supports and restraints not installed prior to the start of the test, as evidenced oy the TRT's review of TDRs 680, 722, 746, 747, 837, 1006, 1032, 1243, 1244, 1665, 1674, 1724, 1786,
- 1799, 1851, 2034, 2106, 635, 709, and 732. Additionally, TPD-1, issued j
against 1CP-PT-34-05, " Steam Generator Narrow Range Level Verifica-tion," identified that a substitution was made for steam generator l water level detectors. TPD-2, issued against ICP-PT-22-01, " Process j Sampling," identified that three radiation monitors were not installed
- ' at the time of the test and stated that they were not needed to meet the test objectives.
~
I In every case reviewed by the TRT, missing components and equipment j were identified and documented in the completed test record. Any outstanding testing which remained because components and equipment l were not installed at the time of the test was tracked by a deferred preoperational testing program schedule implemented by STA-805, ] Revision 0, " Deferred Preoperational To:: ting." STA-805 is a CPSES administrative procedure. In interviews with TUEC personnel, the j j TRT determined that the decision to proceed with the HFT despite missing equipment was made to minimize the economic impact of delay-l j ing the testing program and was deliberated on and concurred in by
- senior TUEC management, the architect-engineer,'and the nuclear I steam system supplier.
The TRT also reviewe'd a master data base computer printout of work items requiring thermally hot plant conditions in order to retest. ! As alleged, there were modifications (about 74), most of which were l i on hangers, snubbers, and other pipe supports, that required thermally
. hot plant conditions, such as during the HFT, for valid ratesting.
.i Thus, the TRT found that while some components and equipment were { not installed during the initial (1983) HFT, they were documented ' l and tracked to be included in the deferred preoperational testing. l (2) monitor In assessing the allegation that TUEC does not intend to check or some components and systems until " heat-up to hot standby" l- or "during power ascension," the TRJ reviewed Integrated Plant l-
- - Operating Procedure IPO-001A, " Plant Startup From Cold Stutdown to J-70
Hot Standby." This procedure specifies that the plant be taken to normal operating pressure and temperature using reactor coolant pumps (not the reactor) as the heat source. This is what was done during the initial HFT. However, it should be noted that some preoperational tests can be done only after fuel loading because the reactor core must be installed to conduct a valid test. Examples of these are: ISU-022A, "RCS Boundary Pressure Test and Leakage"; 150-0228, "Incore Moveable Detector System Alignment"; ISU-021A, Pressurizer Spray & Heater. Capacity Test"; and ISU-228A, Control Rod Drive Mechanism Operational Test." At the time of its review, the TRT learned that IUEC had plans to conduct tests on components and equipment not installed during the initial HFT and tests which require the reactor core to be in place, after fuel load, but before the reactor was placed into operation. However, TUEC now plans to complete those tests, which do not require the reactor core to be installed, prior to fuel loading as sufficient time is now available. The results of those tests and the tests which require the reactor core to be in place must be found to be satisfactory prior to initial reactor cri-ticality. The TRT also learned that there are no HFT items scheduled to occur "during power ascension" except those that require more heat input than can be obtained by the use of reactor coolant pumps alone. For example, steam and feed water piping does not achieve design temperatures until there is sufficient flow, which only occurs at a power level of 25-30L In order to attain this power level, heat input from the reactor is required. Accordingly, this testing cannot be completed until the reactor is made critical and that power level is attained. Section 14.2 of the FSAR and Regulatory Guide 1.68 specify those tests which are to be conducted during power ascension. (3) It is alleged that neither TUEC nor the NRC Region IV staff noticed that major components or equipment were not installed prior to HFT and failed to keep the ASLB informed of the problems encountered. The TRT reviewed HFT-related TDRs and the master data base to determine whether TUEC had documented all outstanding work on the master data base for the Lead Startup Engineer to review prior to each test and that components not installed at the time of testing, but needed for eventual system operation, were documented on TDRs or TPDs, as required by CPSES administrative procedures. For example, as discussed in paragraph 4(1) above, there were 20 TDRs identifying the missing hangers and supports associated with ICP-PT-55-11,
" Thermal Expansion." Each was initiated by the Startup Group and evaluated by TUEC engineering for its impact on the test results.
TUEC performed calculations and installed temporary supports and weights during the test so t>at installed supports, which in normal operation would interact with missing supports, would not yield erroneous data. The TRT also determined that the reason there was no documentation in NRC Inspection Reports to indicate that the Region IV staff was aware of missing components was because the missing components were docu-mented and tracked in accordance with the TUEC administrative proce-dures which provide for such possibilities, and because they were J-71
included in planned and documented future testing activities, i.e., in the deferred preoperational tests. It is not unusual for an applicant for an NRC operating license to defer certain equipment installation in order to proceed with HFT. However, the NRC routine inspection program verifies, before the fact, that a viable system exists to document and track such missing equipment and to ensure that the equipment is satisfactorily tested when it is finally installed, and during testing, that the system is being implemented. This w'.s done by NRC's Region IV staff during various routine inspec-tions of TUEC administrative procedures and was confirmed by the TRT during its review, as described in the preceding sections. It is also alleged that TUEC and NRC Region IV did not keep the ASLB informed of problems encountered during the HFT. Prior to and during its review, the TRT found no instances involving the testing program where ASLB notification by the NRC staff should have been provided and was not. The matter of TUEC not keeping the ASLB informed was raised by CASE directly to the ASLB and is properly a matter for the ASL8 to decide. (4) It is alleged that 60 percent of the test points of 1CP-PT-55-11,
" Thermal Expansion," failed the acceptance criteria, that the traceability of the measuring devices was lost because they were not logged with the data, and that TUEC engineering had provided no justification for the "use as is" determination on piping which did not meet expected thermal expansion values.
The TRT staff determined, through discussions with TUEC personnel and by a review of the completed portions of ICP-PT-55-11, that about 28 percent of the test points (referred to by TUEC as " monitor-ing locations") failed the acceptance criteria. TORS were issued to document all test failures so that TUEC could provide corrective actions and establish retest requirements. Additionally, about 12 percent of the monitoring locations were not measured because of missing equipment at the time of the tests; about 7 percent were invalidated because equipment was removed during the test; and about 3 percent were invalidated because of modifications to equipment after the test. Therefore, about 50 percent of the monitoring locations still required measurements after the thermal expansion test was completed. These locations are included in the deferred preoperational tests. Another related allegation was that, although temperatures were taken and logged during the test, the specific measuring device used at each monitoring location was not logged. As a result the calibration of the measuring device could not be traced to the monitored location with the information contained in the test data packages. The TRT found that the completed test data packages did contain the calibra-tion data for the measuring devices used, but as alleged, the devices While could not be traced directly to specific monitoring locations. pursuing this matter, the TRT interviewed TUEC personnel who partici-pated in the testing and found that a test coordinator maintained a J-72 I i L
log which tied the devices to the specific monitoring locations; how-ever, the log was not made a part of the test data package. The TRT pointed out to TUEC that while the direct connection was not required by the test procedure as written, the data must be included as part of the test data package. The TRT's review of representative TDRs, including TDR-853, 854, 855, 1033,1034,1035,1112, and 1113, identifying questionable data or deficiencies revealed no cases where TUEC engineering had not pro-vided back up data and/or calculations supporting a justification for the "use as is" disposition of a TDR. (5) It is alleged that in conducting the HFT, TUEC considered only normal operating conditions and did not consider accident con-ditions, such as loss of-coolant accident (LOCA) or an earthquake. Each applicant for a permit to construct a nuclear power plant must include the principal design criteria for the proposed facility in its application to the NRC. The principal design criteria in 10 CFR 50, Appendix A, establish the necessary design, fabrication, construction, testing and performance requirements for structures, l systems, and components important to safety which provide for reason-able assurance that the facility can be operated without undue risk to the health and safety of the public, including during accident conditions, such as LOCAs and earthquakes. During its review of preoperational test procedures, the TRT found that TUEC tested safety systems with consideration for accident con-ditions to the extent possible by simulating certain parameters such as temperature, pressure, flow, etc., that might be encountered dur-ing an anticipated accident or emergency condition. This method is permitted by NRC RG 1.68 and, therefore, satisfies NRC requirements. (6) It is alleged that TUEC and the NRC Region IV staff were willing to accept HFT results which were deficient. Final acceptance by TUEC of HFT results does not occur until the Joint Test Group (JTG) has conducted its review of the data and approves the completed test data package. In a sample of 17 out of 25 completed HFT data packages, the TRT found four instances in which not all of the test objectives had been met, yet the JTG had com-pleted their review and had approved the test data package. These instances were: (a) Preoperational test procedure 1CP-PT-02-12, " Bus Voltage and Load Survey," intended to demonstrate that during all modes of plant operation, optimum current and voltage will be present at all the buses and associated equipment. After the test was completed, the STE noted in review of test data that the voltages recorded in paragraphs 7.8.2.1 and 7.8.3.1 did not meet the acceptance criteria specified in the test procedure. A test J-73
deficiency report (TDR) was initiated. Subsequent TUEC engineer-ing evaluation of the out-of-tolerance voltage 3 documented in the TDR required that changes to some of the transformer output settings used during the conduct of the test were necessary to bring the In voltages within the originally specified acceptance criteria. accordance with the test procedure, these changes necessitated that some portions of the test be performed again. However, the JTG approved the data package without requiring these por-tions of the test to be performed again. Therefore, the test data package contained invalid data for that test; thus, the test 3 objective had not been met. , (b) Procedure 1CP-PT-34-05, " Steam Generator Narrow Range Level Verification," intended to demonstrate at hot, no-load con-ditions, that the specified narrow range level channels for each steam generator indicate properly at the upper and lower instrument taps and compare properly with each other for 4 actual changes in steam generator water level. The trans-mitters for level detectors 1-LT-517, 518, and 529 were found defective prior to initiation of testing and, thus, temporary equipment was substituted. The test was performed with the After the test, temporary equipment and declared successful.
! the specified transmitters were installed. The Joint Test Group (JTG) approved the completed test package containing data taken with temporary transmitters. The only retest required after installation of the detectors was cold cali-bration (not calibration at hot, no-load conditions); thus, this test objective was not met and no other requirements were imposed by the JTG to monitor performance when the transmitters are placed in service.
(c) Procedure 1CP-PT-55-05, " Pressurizer Level Control," intended to demonstrate the control aspects of the system in conjunction In addition, there with the chemical and volume control system. was a note on page 12 of the procedure that stated, "This test t is provided to verify the capability of the pressurizer level control system to monitor pressurizer level over the range of i installed instrumentation and to observe that all alarm and control functions are operational." A prerequisite condition (paragraph 6.13) required the plant to be in hot standby con-dition. During conduct of pressurizer level indication testing in accordance with the procedure (paragraph 7.1), the System Test Engineer (STE) noted that a level detector (1-LT-461) was registering marginal readings. He documented this and recom-- After the test was mended a calibration check of the detector. completed, this was done, and it was determined that the detector was out of calibration, and attempts to calibrate it were unsuc-cessful. The corrective action was to replace the detector and perform a cold calibration (not calibration in hot Thestandby condi-JTG-approved tion); thus, this test objective was not met. 4, test data package contained level data taken with a detector that subsequently proved to be out of calibration, thereby invalidat-ing the test data and no other requirements were imposed by the l J-74 r
JTG to monitor the performance of the new detector when it was placed in service. (d) Additionally, during the conduct of Procedure ICP-PT-55-05 dis-cussed in (c) above, the speed of the recording chart for the pressurizer level was changed from 2.5 cm/ minute, as required by paragraph 7.2.6c, to 15 cm/ hour. The TRT determined that this was done to avoid running out of chart paper during the test. This deviation from the approved test procedure should h' ave been documented on a TDR even though, in this case, the chart speed was inconsequential since the recorded trace data were not being relied upon to prove any of the system's per-formance features. The TRT discussed these findings with startup management, including the Startup Manager, who is a JTG member. The Startup Manager informed the TRT that with respect to ICP-PT-34-05 and ICP-PT-55-05, the JTG had made a conscious decision not to require hot calibrations on the instruments in question since the accuracy of their calibrations could be determined during a subsequent plant heatup. While the TRT understood this, it pointed out that the JTG had not specified in the I retest requirements that these hot calibration determinations must be made; it only specified a cold calibration. Therefore, there was no mechanism to draw attention to the fact that these instruments had not been operationally tested previously under hot plant conditions. The TRT, therefore, did not consi' der the test objectives to have been fully met. With respect to ICP-PT-02-12, when the TRT identified the need to perform some portions of the test again as a result of the actions taken to implement TUEC's engineering evaluation of the out-of-tolerance group. voltages, a TDR was immediately initiated by the startup The need for performing portions of the test again was appar-ently overlooked by the JTG during its review. The TRT, therefore, considered that the test objectives had not been fully satisfied and that the JTG review of this data package had been less than adequate. With respect to the alleged acceptance of deficient test results by the NRC Region IV staff, when the TRT review began, the Region IV staff had not yet begun their inspections of HFT-completed test packages. This NRC inspection effort has as an objective to assure that all test data are either within pre-viously established acceptance criteria, or that deviations are properly documented, evaluated and dispositioned. Since Region IV inspection had not yet begun, the implication that the Region IV staff was willing to accept deficient results was not appropriate. Thus, there is no support for the assertion of a lack of candor on the part of the NRC Region IV staff or for the assertion that the ASLB cannot rely on the NRC staff to monitor plant testing. With respect to TUEC, the TRT Test Program Group's findings, dis-cussed elsewhere in this SSER, indicate that the problems identified during HFT were, in general, appropriately and clearly documented and tracked for resolution, in accordance with TUEC administrative pro-cedures developed for those purposes. The TRT review found that J-75
I documentation in the startup group was maintained in an orderly, systematic and readily retrievable manner. Additionally, in its review the TRT Test Program Group interviewed and met with startup_ personnel, including the Manager of Startup, lead startup engineers, startup test engineers and others involved in the testing program. The TRT did not discern any hesitation, lack of knowledge concerning responsibilities, or lack of candor on the part of those personnel,
' nor did the TRT identify any conflicting statements among those inter-4 viewed. Additionally, the TRT conducted a random sample of current startup personne! qualification records and found that the personnel possessed the necessary background and experience to carry out.the responsibilities of their positions. The TRT found no indication of a lack of candor on the part of TUEC startup personnel.
2 The TRT's review of the overall HFT pro-
; 5. Conclusion and Staff Positions: ' gram and a sample of 17 out of 25 completed HFT data packages disclosed 4 instances in 3 test data packages where not all of the test objectives had been met, although the JTG had reviewed and approved the completed
- test. These deficiencies were not part of any specific allegation.
i However, the TRT considers them to be oversights on the part of the JTG Therefore, which raised concerns regarding their review / approval process. the matter is considered to have potential generic implications and to require follow-up action by TUEC. j With regard to the specific allegations, the HFT portion of the pre-j operational test program was found to be comprehensive and, in general,
' conducted with adequate administrative controls and test procedures.
Although the HFT was incomplete, TUEC's plan to complete it after fuel Ioading and prior to initial criticality appeared technically sound and without any safety implications. Subsequently, TUEC altered these plans l and will conduct those tests which can be performed without the reactor core installed prior to fuel loading, since time is now available. l The TRT found no instances involving the testing program when the NRC staff should have provided notification to the ASLB. With respect to TUEC's i notification to the ASLB of problems encountered during the HFT, since CASE-l raised this directly to the ASLB, it is properly a matter'for the ASLB to decide. While problems were encountered during the thermal expansion test, the TRT found that they had been properly documented in accordance with j i administrative controls established for that purpose. The TRT also found
~
that TUEC had tested with consideration of accident conditions The TRT found notosupport the for extent possible as required by NRC guidance. i the assertion that the NRC Region IV staff was willing to accept deficient
- HFT results since they had not yet begun their review. And, while TUEC's I JTG, in the opinion of the TRT, approved two test data packages without imposing appropriate measures to ensure that certain instrumentation was
' accurately calibrated and properly functioning before the plant is made operational, and approved one test data package without recognizing the need to perform portions of the test again, the TRT did not consider these j to indicate a willingness on the part of TUEC to accept deficient test results. There was no evidence found that either TUEC or the NRC Region IV staff was willing to accept deficient test results or that either had ex-It appeared I hibited a lack of candor in identifying problems during HFT. i that the overall objectives of the CPSES Unit 1 preoperational test program I i
- J-76
_ _ _ ~ _ . - - - - - _ .,_. _- _ _ - . . - - _. _ _ - - - - _ _ - - _ _ . - - '
were being satisfactorily met, thus providing reasonable assurance that the plant is properly designed and constructed and that its operation will not pose a threat to public health and safety. While some of the allegations had valid bases, none was considered to have safety significance or generic implications. The findings and conclusions of the TRT with regard to these allegations were presented to the intervenor (CASE) in a meeting on November 7, 1984. CASE had no comments at that time but requested time to review the - transcript of that meeting and to provide any comments thereafter. The TRT agreed to their request. A portion of the allegation discussed in - paragraph 4(4) of this SSER was brought forward by a confidential source. The alleger was not available to discuss the TRT's findings and conclusions.
- 6. Action Required:
- a. Section 4(6) of this report refers to three preoperational tests con-ducted during HFT that the TRT determined were not completed to the extent required by the objectives stated in the test procedures.
Accordingly, TUEC shall review all complete preoperational test data packages to ensure there are no other instances where test objectives were not met, or prerequisite conditions were not satisfied. The four items identified by the TRT staff shall be addressed, with appropriate resolution, in the deferred preoperational tests.
- b. TUEC has informed the TRT that the Station Operation Review Committee (SORC) will review deferred preoperational test data. Since the review of data obtained from the deferred preoperational tests is a function of the 50RC, TUEC shall amend the FSAR to reflect their commitment to the TRT that the SORC and not the JTG will perform these reviews. This requirement, not included in the Sept. 18, 1984, letter to TUEC, is necessary because the current version of the FSAR states that the JTG is responsible for reviewing preoperational test data.
- c. The TRT determined, as indicated in 4(4) of this report, that ICP-PT-55-11 " Thermal Expansion," did not include information needed to trace the measuring devices to the monitored locations, although the information was available in a log maintained by TUEC. TUEC shall incorporate the information contained in the log into the official 1CP-PT-55-11 data package so that the traceability is maintained, and shall also establish administrative controls to assure appropriate test and measuring equipment traceability during future testing and plant operation. .
l l J-77
L Allegati_or, Category: Test Program 2 Unit 2 Test Program
- 2. Allegation Number: AT-12
- 3. Characterization: in support of a proposed contention (No. 26), the intervenor, Citizens Association for Sound Energy (CASE), alleges that un-less ordered to do sa by the NRC Atomic Safety and Licensing Board (ASLB),
Texas Utilities Electric Company (TUEC) will not conduct a testing program on Unit 2, but will rely instead on the results of the Unit 1 testing program to support Unit 2 operation.
- 4. Assessment of Safety Significance: The implied safety significance of this allegation is that safety-related structures, systems, and components associated with Unit 2 would not undergo a testing program to verify that the plant has been properly designed and constructed to assure public health and safety.
The NRC Technical Review Team (TRT) reviewed TUEC's preoperational testing program for Comanche Peak, Unit 2. The TRT also reviewed TUEC's Final Safety Analysis Report (FSAR), Chapter 14.0, " Initial Test Pro-gram," and found it to be consistent with NRC Ragulatory Guide (RG) 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," and RG 1.68, " Initial Test Programs for Water Ccoled Nuclear Power Plants." TUEC is committed in the FSAR to meeting both of these regulatory guides. Regulatory Guide 1.68 requires that all structures, syetems, and components that are important to safety be tested. The Comanche Peak FSAR Chapter 14.2.1, " Summary of Test Programs and Objectives," states that the purpose of the startup program for Comanche Peak Steam Electric Station (CPSES) is to assure that the installed station structures, systems, and components will be subjected to tests to verify that the plant has been properly designed and constructed and is - ready of to operate in a manner that will not ondanger the health and safety the public. The FSAR for Comanche Peak encompasses both Unit 1 and Unit 2. Figure 14.2-3, "Preoperational Test Schedule," and Figure 14.2-4," Initial Startup Test Schedule," indicate that the respective schedules are applicable to both Unit 1 and Unit 2. Accordingly, this statement does not imply that testing will be conducted only on Unit 1. The TRT also reviewed TUEC's "AT/PT Test Index with Schedule, Unit 2-CPSES," (July 18,1984). This document provided an index of acceptance tests (ATs) and preoperational tests (pts), including test numbers, revision numbers, and procedure titles for the projected Unit 2' testing program. :Due to the uncertainty of when Unit'2 construction would be completed, this' document did not show a projected schedule for testing. The TRT compared the Unit 2 index with RG 1.68 and with the Unit 1 index and found them to be consistent. Only systems which are shared by Unit I and Unit 2 and were fully and successfully tested during the Unit 1 testing program were not l scheduled to be tested during the Unit 2 testing program. Examples of l
" shared" system tests (which were listed on "AT/PT Test Index with Schedule, Unit 1 and Common," dated July 9, 1984) included: Waste Gas Sys-tem Leak Check; Control Room Heating and Ventilation System; Telephone and J-79
Radio Systems; Primary Plant Ventilation System; and Primary Plant Ventila-tion Supply System Cooling. The control room heating and ventilation system is typical of the commonality of these shared systems. Units 1 and 2 share
- the same control room, which has one heating and ventilation system.
Because the heating and ventilation system was tested satisfactorily when Unit 1 testing occurred, it need not be tested during the Unit 2 test program. In fact, if Unit 1 is operational, this system will already be in operation when Unit 2 testing takes place, as will be true for the other shared systems.
- 5. Conclusion and Staff Positions: The TRT concludes that this allegation is without basis. TUEC has committed to the NRC staff that Unit 2 would undergo a test program subject to NRC requirements and the TRT confirmed that a Unit 2 test program is planned. Accordingly, this allegation has neither safety significance nor generic implications.
The findings and conclusions of the TRT with regard to this allegation were presented to the intervenor, CASE, in a meeting on November 7, 1984. CASE had no comments at that time, but requested time to review the The transcript of that meeting and to provide any comments thereafter. TRT agreed to their request.
- 6. Action Required: None.
J-80
- 1. Allegation Category: Test Program 3, CILRT
- 2. Allegation Number: AT-7
- 3. Characterization: In support of a proposed contention (No. 26), the intervenor, Citizens Association for Sound Energy (CASE), alleges that the leaks encountered during the containment integrated leak rate test (CILRT) were numerous and of such magnitude that they would have to be corrected and the test repeated before fuel loading.
- 4. Assessment of Safety Significance: The implied significance of this allegation is that the containment building might not be capable of meeting its intended safety function of acting as the final barrier against the release of significant amounts of radioactive fission products to the environment in the event of an accident unless the CILRT was per-formed again with no leaks detected.
A condition for an operating license for a water-cooled power reactor, such as Comanche Peak Unit 1, is that the primary reactor containment building meets the leakage test requirements set forth in 10 CFR 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Reactors." Appendix J of 10 CFR 50 requires preoperational testing of the overall leak tightness of the containment building (CILRT or Type A test) and establishes acceptance criteria for the test. The testing is conducted to assure that total leakage through all designated penetrations and building flaws, if any, does not exceed the value specified in Appendix J or the CPSES Technical Specifications (which are currently under review by the NRC as part of the operating license review process). Both 10 CFR 50, Appendix J, and the Comanche Peak Steam Electric Station Final Safety Analysis Report (CPSES/FSAR), Amendment 12, October 8, 1980, specify the use of the American. National Standard Institute (ANSI) N45.4-1972,
" Leakage Rate Testing of Containment Structures of Nuclear Reactors,"
March 16, 1972, to carry out the test. A later revision of the ANSI standard (ANSI /ANS 56.8, " Containment System Leakage Testing Requirements") prescribes essentially the same test procedure for the CILRT as ANSI N45.4-1972, but prescribes another method for calculating the leakage rate. ANSI /ANS 56.8 has not been endorsed by NRC and is not prescribed in 10 CFR 50, Appendix J. The TRT reviewed the as performed CILRT procedure, 1-CP-PT-75-02, "Struc-tural Integrity Test and Integrated Leak Rate Test," Revision 0 and the resultant test data to determine compliance with 10 CFR 50, Appendix J and the proposed Technical Specifications. The TRT determined that, as alleged, numerous leaks were detected during the first two of three attempts to measure the containment building leakage rate. On each of the first two attempts, when it was determined that the leakage rate would exceed the l maximum allowable rate, the test was terminated, the containment pressure reduced to a safe level for entry into the building, and the suspected leaks corrected. Prior to the third attempt, test personnel identified three containment electrical penetrations (E-49, E-62, and E-68) for J-81 l l
which the individual leakage rates were excessive, but for which a method to stop the leakage was not then apparent. These three penetrations were isolated prior to the third attempt and documented on test deficiency reports (TDRs) for later disposition. The result of the third CILRT The CILRT was observed by ' attempt was considered satisfactory by TUEC. 50-445/83-04) two NRC inspectors (reference NRC Region IV Inspection Report l to ascertain whether the test was conducted in accordance with the approved TUEC preoperational test procedure. The NRC inspectors also independently calculated the leakage rate using the method defined in ANS N45.4-1972 and
' Draft 3 of ANSI /ANS 56.8-1981 to determine the validity of TUEC's test results.
Subsequent to the third attempt, the three isolated electrical penetrations were individually leak tested to establish their specific leakage rates prior to repair. The cause of the leakage was identified as improper assembly of the penetration seals. The penetrations were reassembled and individually leak tested again, with satisfactory results. (Four other penetrations that, in accordance with the test procedure, were required to be open in order to conduct the CILRT were also individually leak tested.) The measured leakage rates from the repaired electrical penetra-tions (and the measured leakage rates from the four penetrations used to conduct the test) were added to the measured leakage rate from the CILRT. This addition was insignificant and did not alter the least significant digit in the previous total leakage rate. The total resultant leakage rate 1 was less than the allowed maximum for the containment building under the
- proposed CPSES Technical Specifications and 10 CFR 50, Appendix J.
During the third attempt, test personnel recorded data and calculated con-tainment building leakage rates as prescribed by ANSI N45.4. These leakage rates remained consistently lower than the maximum allowed in 10 CFR 50, However, the Appendix J, and the proposed CPSES Technical Specifications. calculation of the containment leakage rate included in the summary report submitted to the NRC, as required by 10 CFR 50, Appendix J, (" Comanche Peak Steam Electric Reactor Containment Building Unit One Preoperational Integrated Leak Rate Test," 1983, Decket Number 5C-445, Texas Utilities Generating Company and Addendum, July 1983) was performed using the method prescribed by ANSI / ANS 56.8. This value was consistent with the value j calculated by using the method in ASNI N45.4 and confirmed that the con-
- tainment building leakage was less than that allowed by the CPSES Technical Specifications and 10 CFR 50, Appendix J.
- 5. Conclusions and Staff Positions: The TRT determined that numerous leaks were encountered as alleged during the first two attempts to conduct the CILRT, but that these leakage paths were identified and the leakage was stopped prior to the successful completion of the CILRT, with the exception 1
l of three electrical penetrations. The leakage rates from these penetra-l tions (and four penetrations which were needed to conduct the test) were later measured and added to the total leakage rate. The preoperational leakage rate was calculated and found to be lower than the maximum allowed by NRC regulations, a determination verified throughThe independent CILRT was calcula-performed tions by NRC inspectors and confirmed by the TRT. again without encountering numerous leaks and, therefore, the Containment Building proved to be capable of meeting its intended safety function. J-82 __ . _ _ . _ _ _ _ _ _ _ ~ , . _ __ ______ ,_._ ._ __ _ . - ___ ._.
._ = However, the method for calculating the leakage rate, as reported to the NRC, was as prescribed by ANSI /ANS 56.8-1981, which is not consistent with TUEC's FSAR commitment. While this method differs from that prescribed in ANSI N45.4-1972, to which TUEC had committed, because of the stable and consistent data obtained during the test, the leakage rate which resulted from the use of the calculation method in ANSI /ANS 56.8-1981 would be essentially equivalent to the results which would be obtained using the method in ANSI N45.4-1972. However, it is the TRT's position that TUEC should have either used ANSI N45.4-1972, or provided the NRC with justifi-cation for using a calculational method not endorsed by the NRC to report the results of the CILRT. Further, the TRT considers that conducting the CILRT with three electrical penetrations isolated, though technically insignificant with respect to the test results, does not fully meet the intent of the preoperational CILRT and should not have been done without , specific approval of the NRC staff. These matters were forwarded to the NRC Office of Nuclear Reactor Regulation (NRR) for action. NRR has re-quested additional information from TUEC, identified as FSAR ques-tion QO22.22. In a letter dated December 21, 1984, TUEC responded and submitted appropriate changes to the FSAR text which will be a part of Amendment 54 of the CPSES FSAR. On January 17, 1985, NRR concluded that these matters were resolved as reflected in Item (36) in Section 1.7 of Comanche Peak SSER 6. While these were not safety significant in this case, the deviation from an FSAR commitment, made without identifying it to the NRC, could be indicative of a generic weakness, if other deviations occurred and were not documented and reported to NRC. The findings and conclusions of the TRT with regard to this allegation (with exception of the final NRR disposition noted above) were presented to CASE in a meeting on November 7, 1984. CASE had no comments at that time, but requested time to review the transcript of that meeting and to provide any comments thereafter. The TRT agreed to their request.
- 6. Action Required: Prior to fuel loading, TUEC shall identify all other deviations to the NRC.
from FSAR commitments which have not been identified previously J-83
l i l
- 1. Allegation Category: Test Program 4, Prerequisite Testing
- 2. Allegation Number: AT-14
- 3. Characterization: It is alleged that: (1) prerequisite testing was performed by craft personnel not qualified in accordance with ANSI N45.2.6, " Qualification of Inspection, Examination, and Testing Personnel for Nuclear. Power Plants"; (2) System Test Engineers (STEs) were signing for tests that were conducted by craft personnel when in the majority of cases the STEs were not present during testing; and (3) test documentation.
was made to look as if the tests were performed by STEs, when in fact ; tests were performed by craft personnel and the STEs only reviewed the data. (An allegation similar to (1) and (2) above was also evaluated by the QA/QC Group under QA/QC Category 4, where it was identified as allega-tion QA-91.)
- 4. Assessment of Safety Significance:
The implied safety significance of these allegations is that if prerequisite testing activities were per-formed by craft personnel not trained and qualified in accordance with industry standards endorsed by NRC, errors could be made which could , affect the prerequisite test results. The CPSES Final Safety Analysis Report (FSAR), Section 14.2, describes the initial test program. It was implemented by Texas Utilities Electric Company (TUEC) through a series of administrative procedures contained in the "Startup Administrative Procedure Manual." The CPSES initial test program is divided into three successive phases: (1) prerequisite test-ing, (2) preoperational testing (which occurs prior to fuel load), and (3) initial startup testing (which occurs after an operating license that permits fuel load is issued by the NRC). These allegations address the preoperating license category of " prerequisite" testing, the first phase in the initial test program. " Prerequisite" testing is performed to verify the complete installation, cleanliness, and initial operability of indivi-dual plant components and is also referred to as initial checkout. . Testing in this phase is conducted using a series of generic instructions contained in the TUEC "Startup Prerequisite Test Instruction Manual" and involves checks of such things as electrical resistance, transformer polarity, relay and circuit breaker operability, motor rotation and initial operation, initial pump operability, systems cleanliness, and piping support adjust-ments.
"Preoperational" testing follows the " prerequisite" testing phase and is conducted prior to fuel loading to demonstrate the capability of components, systems, or structures to meet safety-related performance re- ' quirements as stated in the FSAR and as accepted by the NRC. These tests can only be conducted and supervised by personnel who are qualified to ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel."
i-In assessing the allegation that prerequisite testing is being performed by craft personnel who do not meet the qualification standards of ANSI N45.2.6, J-85
_ _ _ _ _ ~ _ . _ _ _ _ _ __. _ . _ _ _ _. __ i l l i
" Qualification of Inspection, Examination, and Testing Personnel for Nuclear Power Plants," the TRT found that TUEC was using craft personnel who do not qualify as " test personnel" under that standard to assist with prerequisite testing. The TRT interviewed TUEC management representatives who stated that STEs are permitted to use qualified craft personnel to assist with prerequisite testing. TUEC's position was that craft personnel ' who support testing are not required by ANSI N45.2.6 to be qualified as test personnel. The TRT reviewed TUEC's FSAR, ANSI N45.2.6 and NRC RG 1,58
(" Qualification of Nuclear Power Plant Inspection, Examination and Testing Personnel") and, in particular, Regulatory Position C.7 of that guide. The TRT found that it permitted the use of craft personnel for data-taking and equipment operation provided they are supervised by a qualified individual and that they have sufficient training to assure an acceptable level of performance. The TRT determined from review of selected pre-i requisite test instructions that the tests involved work normally within the expertise of journeyman level craft personnel. The TRT reviewed the personnel records of craft personnel who are used to assist with prerequi-site testing and found that they were generally at the journeyman level in their crafts. They were also receiving indoctrination in the testing work; e.g., the Electrical Test Group (ETG) craftsmen were required to read and understand 10 pertinent startup administrative procedures and 14 pre- !' requisite test instructions. The TRT interviewed STEs, ETG craft persons, and the ETG foreman. In addition, one of the TRT members, who has been assigned as a NRC Resident Inspector at CPSES since December 1983, period-ically observed ETG craft persons at work in the field assisting in test-ing activities. No apparent qualification deficiencies were found for the type of work they were performing and, in some instances observed, the i ETGs' knowledge of the components and test equipment directly contributed to the successful completion of the test. 1 The TRT reviewed TUEC Administrative Procedure CP-SAP-21, " Conduct of Testing" to determine the administrative controls established for the use of craft personnel to assist with prerequisite testing. The TRT found that in all cases with prerequisite tests, the test engineer, usually an STE, first ensures that all conditions required to proceed with the test are satisfied, as stated in the test instruction. The STE must indicate this by signing that step in the test procedure. If an STE assigns a craft person to assist with a portion of the prerequisite test, the STE must first assure that the craft person is adequately experienced to do the work by having directly observed him in that activity. When a craft person is used to measure and record data, that person must sign for the data he/she has recorded in the same manner as if it had been recorded by an STE. Since the STE is directly responsible for the proper conduct of the test, the STE must evaluate the completed test and resulting data against the test acceptance criteria to determine if it is satisfactory. The STE signs the test data sheet to indicate the satisfactory completion of that review. The TRT considered that TUEC's practice of utilizing craft personnel to assist with prerequisite testing to be consistent with the applicable industry guides and standards and in conformance with the FSAR commitment. i J-86 i
. . , _ . _ , . . . . . _ _ , , _ _ , , _ _ , . _ _ _ _ , -_m.,._ ,. , ,__ _ , , _ . _ . _ .- , __
The allegation also implied that STEs were signing for testing by craft personnel when the STEs were not present during the testing. The TRT determined that in some instances this could occur, i.e., while a craft person was measuring and recording data he may not have been directly observed by the STE. However, since an STE has to initiate a test and execute steps in the test which involve equipment operation, an STE has to be present for the test to proceed. In light of the experience level of the craf t personnel, the nature of the work they performed, the adminis-trative controls established by CP-SAP-21 and the responsibilities of the STEs for proper completion of the test, including test data review, the TRT did not consider that continual observation of craft personnel engaged in prerequisite testing was warranted since it appeared to fall within the generally accepted definition of a supervised activity and is not required by applicable industry standards and guides. It was also alleged that documentation of prerequisite testing can mislead a person into believing that an STE conducted the test when, in fact, it was performed by craft personnel. The TRT's review of 35 test data pack-ages and interviews with startup personnel confirmed that craft personnel, when used to measure and record data on a prerequisite test data sheet, did sign for the data they recorded. The craft persons' signatures on the data sheets clearly indicated that they had recorCed the data; however, the STE was held responsible for the satisfactory completion of the test, evaluation of the resultant data, and for signing the dats sheets. The data sheets were also signed by a test engineer with higher qualifications than the STE to indicate his review of the recorded data. This practice is in accordance with TUEC's Administrative Procedure, CP-SAP-21. However, in its review of the 35 test data packages, the TRT review found that craft personnel verified and signed for initial conditions on some pre-requisite test data sheets, contrary to Section 4.10.9 of CP-SAP-21 " Conduct of Testing," which requires that this be done by the STE. Further investigation revealed a memorandum issued by the Lead Startup Engineer on March 31, 1983, countermanding this requirement of CP-SAP-21. The subject of the memorandum (STM-83084) was "ETG Personnel Schedule Change," but it also indicated that craft personnel (ETG) may verify prerequisite conditions for Prerequisite Test Instructions XCP-EE-1 and XCP-EE-14. Issuing such a l memorandum in lie': of executing a properly approved change to CP-SAP-21 is ) in violation of CP-5AP-1, "Startup Administrative Procedures Manual," I Section 4.4.3.1, which requires a permanent or interim change to be approved and issued to all manual holders in accordance with CP-SAP-1. It appears that as a result of the memorandum, 24 of the 35 tests reviewed by the TRT had prerequisite conditions improperly verified by craft support personnel. Fifteen were XCP-EE-14, but nine were XCP-EE-24, " Fixed Battery Pack Operated Emergency Lighting Units," which were not authorized by the memorandum. J-87
i
- 5. Conclusion and Staff Positions: As alleged, TUEC utilized craft personnel who were not qualified to ANSI N45.2.6 standards to assist with prerequi-site testing activities. While qualifying craft personnel to that stan-dard would be more conservative, the method utilized by TUEC is permitted by ANSI N45.2.6, as augmented by NRC Regulatory Guide 1.58 (Regulatory Position 7), which permits personnel who do not meet ANSI N45.2.6 to engage in data-taking and equipment operation provided they are supervised by a qualified individual and that theyBased have on sufficient knowledge its review, the TRT to foundensure an acceptable level of performance.
that the craft personnel used to assist with prerequisite test activities were appropriately indoctrinated in the administrative and prerequisite test procedures applicable to their work, performed the work under STE supervision, and performed work that was within the journeyman level of expertise. While they may not have been under the constant supervision of an STE, as the allegation implies, this is not required by ANSI N45.2.6 or Regulatory Guide 1.58. The TRT considers that because of the relatively routine nature of the work, and because the prerequisite test'results were reviewed and evaluated against the acceptance criteria by the STE respon-sible for the test, and were subsequently reviewed by a test engineer with higher qualification than the STE, adequate technical supervision and over-sight were being exercised. The TRT did not find, as alleged, that the test documentation was made to look as if an STE performed the test when, in fact, it had been performed by a craft person. The TRT found that when craft personnel took and recorded test data, they signed the entry, and the STE's signature on the data sheet only indicated that the resultant data had been evaluated against the acceptance criteria by the STE and was found to be satisfactory. This practice is consistent with the TUEC pro-cedure CP-SAP-21 which directs the conduct of testing activities. This procedure was widely disseminated onsite and was contained in TUEC's sys-tem of manuals and procedures. Therefore, the TRT concludes that Accord-the practice is not misleading and, as implemented, is satisfactory. ingly, this allegation has neither safety significance nor generic implications. However, the results of the evaluation pertaining to inadequate qualifi-cations of preoperational test personnel will be further assessed as part of the overall programmatic review concerning procedures addressed under QA/QC Category 4.* The alleger This allegation was brought forward by a confidential source. was not available to discuss the TRT's findings and conclusions.
*The TRT evalsation of QA/QC allegations is in progress and will be published in a subsequent supplement to this SSER.
J-88
i
- 6. Action Required: TUEC shall rescind memorandum STM-83084 of March 31, 1983, which was issued in conflict with CP-SAP-21, and take action to ensure that there are no other memoranda issued which conflict with -
approved procedures. TUEC shall also conduct a review of all other prerequisite test records to determine those that had prerequisites signed by craft personnel, and. assess the impact of those improper-verifications on subsequent testing. e 1 I r l l I i l i 1 J-89 3 1
- , , , . < , , , . . . . , , , , - a , , - - , - , , . ..,: -
Test Program 5, Preoperational Testing
- 1. Allegation Category:
- 2. Allegation Number: AT-15 It is alleged that the preoperational test program l/ is
- 3. Characterization:
flawed because (1) several system test engineers STE(STEs) may test for electrica ility mechanical plant systems may work on the same system, or one f in a part of many systems, a condition causing " which conf and overlaps, (3) STEs are not provided with a " computer for printout informs them of all tests required on a system, (4) calculationst instantaneous trip settings on approximately 100 b Final Safety Analysis Report (FSAR) commitments, (6) ih a t system ever can pa through both the prerequisite test and the preoperationalwith ided test w t ou undergoing an energized functional test, and (7) ST researching and validating drawings. The implied safety significance of d in the FSAR
- 4. Assessment of Safety Significance:thesei allegations is that if safety were not properly tested, there would not be suffic en syste'ms will meet their inteaded safety functions in service.
The TRT reviewed the " prerequisite" testing method Through aused series by TUEC EE-8, that the systems were ready for preoperational testing. i " of generic tests, such as XCP-EE-1, "Megger Hi/ Pot Testing," XC
" Control Circuit Functional Testing," or XCP-ME-1, " Initial Pump TUEC verified that construction was completed as required i The in orde structures, systems, and components to undergo preoperational t cali- test prerequisite testing phase included such tests asl tests initial d separa-of instrumen bration, system piping flushes and cleaning, wiring i continuity an tion checks, hydrostatic pressure tests, and initial funct ona Prerequisite testing is discussed further in Test Program components.
Category 4. These tests facilitate the safe and orderly progression68, and as to th preoperational testing phase, as outlined in Regulatory Guide 1. i whether structu committed to in the CPSES FSAR, which determines systems, and components meet their safety-related design funct o In assessing the allegation that STE assignmentsment areSTEresponsibl fusion and possible omissiens, the TRT discussed TUEC's per- proc plant system assignments dering an interview with startup manag sonnel. The TRT found that each STE was assigned E designated as by the star leader to a system or subsystem. subsystems and had two or three STEs i assigned, ts" sheet, with the leader. ible for each ground and experience and were' documented onsystem. a " System Prior to As' which the STEs used to keep track of who was respons System assignment sheets have been in use t since System / about mid The TRT that time, STE system assignments were 80, which h t this docum ! l reviewed a Master System / Subsystem Index
~
infornation was available at least since that date. r J-91 I l l l
In interviews with startup personnel which included STEs, the TRT found no indication of confusion or gaps between systems and subsystems because of the STE system assignment process. As documented in CP-SAP-2, "Startup Program Organization and Responsibilities," and CP-SAP-21, " Conduct of Testing," the STEs are responsible for ensuring that their assigned systems are properly tested and that their tests are coordinated The STEswith wereother STEs also respon-responsible for interconnecting system tests. sible for cooperating with other STEs when the scope The of testing same overlapped practice of subsystems under their respective responsibilities. cooperation is implemented in cases where startup work authorizations (SWAs) applied to more than one system or subsystem. An SWA is documentation of work which is required on structures, systems, and components under the custody of the startup group. The master data base, a multi-functional computerized tracking system initiated in May 1983, listed Prior to May and 1983, trackeda outstanding work and deficiencies on subsystems. similar, manual listing and tracking system for outstanding work and deficiencies, called the Master System? Punch List, provided for this. The TRT considered that the STEs had adequate information and administra-tive controls to preclude confusion among STEs regarding system assign-ments and that adequate administrative controls had been established to avoid omissions. The allegation also implied that the numbering system used to identify sub-systems was a " dual" system which caused confusion. The TRT determined that one component could appear on two interconnecting system or subsystem An diagrams if it happened to be on a boundary between the two systems. example of such a " dual" numbered component could be a motor-operated valve. The valve appears on the fluid system diagram under the fluid system de-signation number, while the motor and control circuits appear on the elec-trical system diagram under the electrical system designation number. In most cases, different STEs are assigned to the two systems; thus, the valve would be included in the testing of both systems, creating an overlap. The TRT does not consider this practice to be confusing, but rather conservative, since the component is tested twice. But if an electrical work item (SWA) was generated on such a component and it was erroneously assigned to the STE who had responsibility for the fluid side of the component, it would be necessary for the STE who was assigned in error to coordinate with the other STE to ensure the work item was followed to completion and the com-ponent tested again as required. Based on interviews, the TRT found that this degree of cooperation was common among STEs and did not cause problems. Additionally, the TRT found that the master data base system, in conjunc-tion with other administrative controls, ensured that open work items and retesting would be completed as required. The TRT found no indication of confusion or gaps (missed tests) between systems and subsystems because of the " dual" numbering system. As alleged, the TRT found that the STEs are not initially provided with a At other plants, computer printout of testing that is required on a system. an index of required tests may be provided as a package by a contractor. As part of However, at CPSES, the startup program was undertaken by TUEC. TUEC's program, the STE is responsible for making an initial determination of what testing is required using design specifications, drawings, the FSAR, and other applicable documents. When the STE has made this determination, I the startup group leader and the Joint Test Group (JTG) review it for J-92
completeness. When it is complete, a test index is published which lists the number and name of the tests and, after testing begins, the status of all tests. The test index is routinely issued to the group leaders and is available to all STEs. The TRT considers this practice to be acceptable. In its investigation of the allegation related to incorrect instantaneous breaker trip settings, the TRT found that instantaneous trip points for 74 miscellaneous circuit breakers had been set at the specific values called for in the design drawings provided by the plant's architect-engineer (A/E) early in prerequisite testing. None of these 74 circuit breakers were safety related. As a result of a few " nuisance" trips, i.e., some breakers instantaneously tripping at locked rotor current values, TUEC startup engineering contacted the A/E concerning that problem, and in September of 1980, this resulted in a revision to XCP-EE-14, " Molded Case Circuit Breaker and Thermal Overload Relay / Heater Testing," which is the generic procedure for testing circuit breakers. The revision incor-porated a formula for determining the correct trip settings by actual calculations. The calculation method includes the motor starting kVA, horsepower, voltage, and full-load current; factors which may not have l been known precisely when the setpoints were previously calculated by the A/E. Therefore, using the actual factors from motor name plate data resulted in a more accurate calculation for the setpoint for a particular circuit breaker. This allegation stems from a situation on or about i March 15, 1984, when an ETG technician, who was verifying data from the results of circuit breaker testing, using XCP-EE-14 in order to establish a computerized data base, found minor variances in trip setpoints for 21 circuit breakers. The technician was using the calculational method of determining the setpoints included in the current revision of XCP-EE-14. He informed startup management of this finding and the startup group pursued the apparent variances. Shortly thereafter, the startup group realized that the reason for the variances was that the trip setpoints for those circuit breakers had not been calculated on site per XCP-EE-14, but had been set in accordance with the trip points provided by the A/E prior to the revision of XCP-EE-14 which incorporated the formula for calculating the trip setpoints. Since the differences were small (within 10%) and the circuit breakers and their associated equipment were not safety related, TUEC did not reset the trip setpoints. The TRT found that the equipment involved included the turbine building roof exhaust fan, cir-culating water traveling screen, polymer mixer, and other similar compo- t nents. The TRT noted that TUEC engineering had appropriately considered the situation, confirmed that no safety-related equipment was affected, l and that the problem did not involve an error in calculations. ; i It was also alleged that some prerequisite testing was not repeated as part of the preoperational testing, and that, therefore, the prerequisite tests were being used to prove FSAR commitments. FSAR Section 14.2.1 states that prerequisite testing is one of the three major phases of the initial test program at CPSES; the other two are preoperational and initial startup testing. Since it is the initial testing phase and is included in the overall program, there is no need to repeat successful prerequisite tests during the preoperational tests. The allegation also implied that a " system" can pass through both prerequi-site and preoperational testing without ever undergoing an energized functional J-93
test and that it was highly probable that it has happened with " light indica-tors." An energized functional test is one which is conducted with elec-trical power supplied to the particular component being tested (primarily control circuits) to ensure that it operates correctly. It is initially done at CPSES during prerequisite testing using procedure XCP-EE-8, " Con-trol Circuit Functional Testing." However, in some instances it may not be possible or practical to provide the component (circuit) with electrical This is power. In such instances, only a continuity test is conducted. permitted by XCP-EE-8. A continuity test ensures that there is an electri-cal conductor (wire) between two specified points in question, but does require that the component be energized. This is generally a sufficient alternative method for initial testing of control circuits and, in parti-cular, for testing indicating light circuits when power is not available. The TRT found that there were cases where some circuits, particularly those for lights indicating valve and breaker positions, may only have had continuity checks during prerequisite testing without having been included as a specific step in the preoperational testing procedure. However, in interviewing startup personnel, the TRT found that even if an indicating light circuit were not energized during prerequisite testing, it is energized during the preoperational testing of the components to which it is ccnnected, since preoperational testing is performed with components in an electri-cally energized condition. When a preoperational test requires a motor-operated valve to be opened, it must be energized, and the operator would expect to see a change in the position indicating light when the position of the valve changes from closed to open. If this did not happen, the operator would indicate that deficiency to the STE so that the cause could be investigated. Most of the time when this happens, it isIf caused not, by a burned out bulb, which the operator replaces on the spot. the STE documents it for resolution. The TRT considered this approach to be reasonable in light of a successful continuity test and the fact that plant operators monitor indicating lights on a routine basis. The TRT could not identify any safety-related circuit which would not be energized during preoperational testing of its associated component in which a deficiency would not be evident. It was also alleged that system drawing packages were being provided to the STEs by the Document Control Center (DCC) with design change authorizations (DCAs) several years old that were not reflected on the current design drawings; that packages were being issued to STEs with DCAs issued against other packages; that print changes were being issued with no DCAs in the packages; and, that there was no procedure to ensure that the STE had current drawings and design information with which to conduct a valid test. The TRT interviewed three STEs-who were responsible for major fluid and electrical systems at CPSES. At each interview, the STE commented that the substance of the allegation relative to outdated design drawings had been true in the past, but that improvements have since been made. That portion of the allegation dealing with the lack of procedures could not be substan-tiated because those STEs interviewed insisted that there were always procedures which charged the STE with the responsibility of ensuring that he had the latest design information. The TRT confirmed that CP-SAP-21,
" Conduct of Testing," Section 4.9, required the STE to use current infor-mation when he was preparing to conduct a test. Thus, the responsibility was placed on the STE to ensure he was working with current information.
To accomplish this, the STE was required to go to the document control J-94~ l
4 1 , center and update the documents. This apparently was very time-consuming and burdensome. The STEs.who were interviewed told the TRT that after much , i discussion with TUEC management, the design information provided to STEs i had improved greatly and, at the present time, only systems such as vents and floor drains continue to be a problem. The STEs are now able to l obtain current drawings from a satellite document control center located t closer to their work station, which .nakes this task less burdensome. During the TRT's review of the Test Program area, it did not find any indication of deficient testing activities which could be attributed to , i
- this problem, either past or present. However, in light of the number and nature of the problems found in the document control system by the TRT QA/QC Group (reference QA/QC Category 5, " Document Control"), the TRT does not consider that there is a sufficient certainty that these document control problems did not affect the testing program. The TRT believes that i
TUEC must provide NRC with assurance that all structures, systems, and com-ponents were appropriately and adequately included in the testing program. Additionally, the TRT believes that TUEC should review the process by which test personnel ensure that the latest design information is used when preparing, reviewing, and approving test procedures, conducting tests, and reviewing and approving test results in an attempt to reduce the heavy reliance on the motivation and initiative of individuals, as is required by the current process.
- 5. Conclusion and Staff Positions: While some of'the allegations were found to have a valid basis, the TRT concludes that the allegations have neither safety significance nor generic implications with the exception of 7, below, the significance of which will be dependent upon the results of
! information that TUEC provides to NRC about how past document control
{. system problems may have affected the testing program. With regard to the specific allegations, the TRT reached the following i conclusions: i (1) The process for STE assignment to systeme did not cause confusion or omissions, and the STEs appeared to be in control of the systems for 1 which they were responsible. t (2) The " dual numbering" system did not cause confusion, but overlaps did occur at system boundaries. These overlaps could only have caused a component to have been tested more than once, which is conservative. i (3) STEs are not provided with a computer printout detailing the testing required on systems to which they are assigned. The STEs are respon-sible for making this determination, and it is reviewed and, when complete, approved by the STE's supervisor and the JTG. The TRT found this acceptable. (4) Calculations, when required, were performed properly for the instantaneous trip settings on circuit breakers, and variances found by an ETG technician were not the result of calculational errors and were of no safety-related consequence. (5) Portions of prerequisite tests are being used to satisfy initial test ~ requirements, but, as stated in the FSAR, prerequisite testing is a ! a a J-95 .
major phase of the initial test program. Therefore, the prerequisite tests, in conjunction with the preoperational tests, satisfy FSAR commitments. (6) Although some electrical circuits, including indicating lights for components may not be specifically subjected to an energized func-tional test, the preoperational test program subjects all systems and components committed to in the FSAR to an energized operating condi-tion as a minimum and, as such, any deficiences would be apparent. (7) No problems were identified by the TRT as a result of the STEs having During to pursue design information updates on their own initiative. the timeframe that the alleged difficulties in obtaining current Care was not being design information occurred, there was a problem. adequately exercised in providing updated packages to the STEs, and the Document Control Center (DCC) was not conveniently accessible to STEs, thus making the STEs' job burdensome. As a result of an upgrade in the document control system in April 1983, satellite DCCs were established to bring necessary information closer to personnel needing it. The TRT determined, through interviews with STEs assigned to fluid and electrical systems, that the problem no longer exists to any degree of significance. However, the TRT believes that TUEC should establish measures which do not rely so heavily on an STE's motivation and initiativt to obtain current design information. Additionally, as a result of problems identified in the document < control system by the TRT QA/QC Group, TUEC shall provide NRC with reasonable assurance that past document control system problems did not adversely affect the testing program.
- 6. Action Required: TUEC shall establish measures to provide greater assur-ance that STEs and other responsible test personnel are provided with cur-rent design documents and change notices. Additionally, TUEC shall provide NRC with reasonable assurance that past document control system problems did not adversely affect the testing program.
J-96
1. Allegation Category: Test Program 6, Lack of Management Conservatism
- 2. Allegation Number: AT-16 3.
Characterization: It is alleged that Texas Utilities Electric Company (TUEC) startup management had a tendency to interpret its commitments to the Final Safety Analysis Report (FSAR), Chapter 14, " Initial Test program," and to applicable NRC Regulatory Guides (RGs) liberally rather than conservatively.
- 4. Assessment of Safety Significance: The implied safety significance of this allegation is that such tendencies could lead to plant testing at a standard below that required by the NRC, which in turn could potentially affect public health and safety.
The primary basis for this allegation appeared to be that the TUEC Start-up Group did not require craft personnel who support testing activities to be qualified to ANSI N45.2.6-1978. The TRT review of that allegation, presented in Test Program Category 4, concluded that, while qualifying those personnel to ANSI N45.2.6 would have reflected a more conservative management attitude, TUEC did not commit to that level of qualification 1.n the FSAR. ANSI N45.2.6-1978, Section 1.2, leaves the imposition of its requirements to the discretion of the employer for personnel who perform work which is well within their normal craft expertise, e.g., calibration and installation checkouts. TUEC exercised its discretion and did not qualify craft personnel who supported the testing activity to ANSI N45.2.6. Additionally, at Comanche Peak Steam Electric Station (CPSES) that work is performed by craft personnel under varying degrees of supervision provided by qualified System Test Engineers (STEs) who are held fully responsible for the correct performance of that work and for the review of data recorded. In order to determine if there were any other bases for this allegation in the test program area, the NRC Technical Review Team (TRT) reviewed FSAR Chapter 14, which describes how the testing program is to be carried out, and the RGs to which TUEC committed. The'se were compared with TUEC's Startup Administrative Procedures and Startup Quality Assurance Plan, which prescribe in detail the conduct of the testing program. In addition, the TRT reviewed procedures related to the test program in Test Program Cate-gories 1 through 5 and 7. With the exception of some minor deficiencies identified in Test Program Categories 1, 3, and 4, the TRT did not find any substantive evidence that the Startup Group interpreted FSAR commitments or RGs in a nonconservative manner. The TRT found, however, that some of the decisions made by startup manage-ment may have appeared to be less than conservative. Through discussions with startup management personnel, the TRT perceived this to be due to the heavy workload and schedule pressures inherent in a testing program of such magnitude. These burdens apparently resulted in decisions by startup management, in the interest of expediency, to delay some parts of a particular test to a later date when the workload and impact on schedule would be lessened. The TRT found severa1 examples of this with respect to preoperational testing. J-97
l 1 i
' l One such example was the TUEC decision to conduct the containment integr-ated leak rate test (CILRT) with three electrical penetrations isolated.
While it was technically reasonable to do that (as long as certain controls were maintained), it is preferred by the NRC that this test be conducted with the Containment Building as close as possible to the configuration it will be in during normal plant operation, i.e., with no penetrations isolated. An allegation concerning how the CILRT was conducted is dis-r ! cussed in detail in Test Program Category 3. 1 Another such example concerned preoperational tests which were originally scheduled to be performed prior to fuel load, but for which TUEC was seeking NRC approval to defer until after fuel load. The Hot Functional i Test, in particular, is discussed in detail in Test Program Category 1.
' These decisions were apparently made because of schedule considerations and, while not the most. conservative course of action, nonetheless were acceptable from the point of plant safety. However, TUEC currently plans l
to perform these tests prior to fuel loading, since additional time is now J available.
- 5. Conclusion and Staff positions: The TRT found no substantive reason to believe that TUEC startup management has a tendency to liberally interpret FSAR commitments and NRC Regulatory Guides in the area of testing. As discussed above, startup management has made decisions which the alleger could have construed as being less than conservative. The TRT found that the administrative controls that TUEC had developed and is implementing for the conduct and surveillance of preoperational testing, are suffi-ciently comprehensive to reveal safety-significant or generic problems.
Accordingly, this allegation has neither safety significance nor generic implications. This allegation was brought forward by a confidential source. The alleger was unavailable to discuss the TRT's findings and conclusions.
- 6. Action Required: None.
i i r 4 J-98 i a
- , . ~ , - . . _ , n-- -,,..,n, r . -n,-+-. , - - - - , w v, - - - - - ,. -n -- - -n ,, ,. . , , , - , . -,..n.
1. Allegation Category: Test Program 7, QA Program for 3tartup Testing Activities is Minimal
- 2. Allegation Number: AT-18
- 3. Characterization: It is alleged that the quality assurance (QA) program for startup testing activities is minimal.
- 4. Assessment of Safety Significance: The implied safety significance of this allegation is that the QA program for testing activities may not have been sufficient to ensure that the testing program met its objective, that is, demonstrated that plant structures, systems, and components were capable of performing their intended safety-related functions.
Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) Appendix B, Criterion XI, requires a testing program to be identified that will demon-strate the satisfactory performance of safety-related structures, systems, and components, and that the testing be conducted in accordance with written procedures which incorporate the requirements and acceptance cri-teria of applicable design documents. Appendix B, Criterion X, requires an inspection program to be established to ensure that activities affecting quality, such as testing of safety-related structures, systems, and compo-nents, are carried out properly. The NRC Technical Review Team (TRT) reviewed the programs that Texas Utilities Electric Company (TUEC) had established and implemented in order to meet these NRC requirements. The TRT's review of the prerequisite and preoperational testing programs is described in Test Program Categories 1 through 6. The TRT reviewed TUEC's QA program for inspection of testing activities. The QA program is described in the Comanche Peak Steam Electric Station (CPSES) Final Safety Analysis Report (FSAR), Chapter 17, and in the CPSES Startup Quality Assurance Plan. This plan delineates responsibilities and measures for accomplishing and controlling testing activities. TUEC's Quality Assurance Manager was responsible for verifying proper implementa-tion of the plan. QA surveillance activities were assigned to the Con-struction and Startup/ Turnover Surveillance (CSTS) Group which was located on the plant site and reported directly to the Quality Assurance Manager. QA audit activities were assigned to the TUEC Corporate QA group. In order to determine the extent of the CSTS Group's surveillance of testing activities, the TRT reviewed CP-QP-19.6, " Surveillance of Con-struction and Startup/ Turnover Activities" and referenced documents which prescribed the method for, and frequency of, conducting surveillances. The TRT found that a surveillance schedule, which was updated monthly to accommodate changes in the testing schedule, dictated the frequency of the QA surveillances by the CSTS group. The schedule was prepared by the CSTS staff and approved by the CSTS Supervisor, as required by CP-QP-19.6. The schedule required surveillance of certain attributes during the conduct of each preoperational test and a minimum of 30 percent of the , prerequisite tests associated with each preoperational test. The pre-requisite test procedures are generic, i.e., the same procedure is used to test each similar component for some basic functional attribute. Prerequisite tests are performed to verify such things as complete installation, functional operability, and cleanliness. Therefore, the J-99
l smaller sample size for surveillance of prerequisite test activities is ! appropriate. Preoperational tests, on the other hand, are not generic, t.e., each procadure is different and is especially prepared to test per-formance characteristics to verify that structures, systems, and compon-ents meet their safety-related design functions. Preoperational tests are the NRC-required performance proof tests. The TRT considered that the surveillance frequency established by TUEC was consistent with general industry practice and was being carried out in accordance with the surveillance schedule. The CSTS surveillance schedule also covered reviews of theThese administrative procedures by which the startup group conducted its program. reviews were scheduled to cover each administrative procedure at least annually. The TRT found that a detailed checklist was prepared by the assigned CSTS surveillance specialist for each test surveillance. These checklists referenced applicable drawings, procedures, and regulatory requirements, and included such attributes as the qualifications of startup personnel, verification of equipment performance characteristics, proper documenta-tion of test results, witnessing of testing activities to verify adherence to procedures, use of correct revisions to applicableAdditionally, testing documents, the TRT and proper completion of prerequisite conditions. noticed that QA " hold points" were designated in these preoperational test procedures, which were reviewed as part of the TRT's review dis-cussed in Test Program Categories 1, 3, 4, and 5. The existence of QA hold points in the test procedures indicates that the CSTS group also performed specific reviews of these test procedures before the start of a particular test, in order to determine which portion needed to be verified by 0A. The TRT reviewed the results of 30 planned surveillances (out of 174) conducted during 1982, 1983, and the first half of 1984, as well as 5 unplanned surveillances (out of 37) conducted during 1983 and the The first half of 1984 and found that they had been adequately implemented. findings indicated that generally thorough surveillances had been conducted. In addition to these surveillances, the TRT reviewed the results of five audits (out of seven) conducted by TUEC's Dallas QA group between late 1982 and the first half of 1984 to determine the extent of involvement by TUEC Corporate QA in the testing program. These audits were found to be comprehensive, and the frequency at which they were conducted was consis-tent with that established by DQP-CS-4, " Procedure to Establish and Apply a System of Pre-Award Evaluations, Audits, and Surveillances." The audits were also commensurate with the safety significance and pace of the pre-operational testing activities discussed in NRC Regulatory Guide 1.33. The TRT concludes that the frequency and
- 5. Conclusions and Staff Position:
degree of TUEC's QA program for testing activities was appropriate, com-mensurate with the safety significance of the specific activity under surveillance, and in compliance with NRC requirements. Accordingly, this allegation has neither safety significance nor generic implications. This allegation was brought forward by a confidential source. The alleger was unavailable to discuss the TRT's findings and conclusions. J-100
1 1
- 6. Action Required: None.
J-101
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.. \ Attachment 3 k -h k
$%,$ a y ,f UNITED STATES NUCLEAR REGULATORY COMMISSION
.....f WASHINGTON, D. C. 20555 Dockets: -
gp g g Texas Utilities Electric Company Attn: M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Daar Mr. Spence:
SUBJECT:
COMANCHE PEAK REVIEW On July 9, 1984, the staff began an intensive onsite effort designed to complete a portion of the reviews necessary for the staff to reach its decision regarding the licensing of Comanche Peak Unit 1. The onsite effort covered a number of areas, including allegations of improper
- construction practices at the facility.
The NRC assembled a Technical Review Team (TRT) responsible for evaluating most of the technical issues at Comanche Peak, including allegations. The TRT has recently identified a number of items that have potential safety implications for which we require additional information. These items are ' listed in the enclosure to this letter. Further background information - regarding these' issues will be published in a Supplement to a Safety Evaluation Report (SSER), which will document the overall TRT's assessment i of the significance of the issues examined. The items in the enclosure to this letter, which are in the general areas of electrical / instrumentation, civil / structural and test programs, cover only a portion of the TRT's effort. The TRT evaluation of items in the areas of , mechanical, QA/QC, and coatings, and its consideration of the programmatic ! implications of these findings, are still is progress. A summary of these issues will be provided to you at a later date. You are requested to submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified. This program plan and its implemen- i tation will be evaluated by the staff before NRC considers the issuance of I an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic implic-ations on safety-related systems, programs, or areas. The collective - significance of these deficiencies should also be addressed. Your program plan should also include the proposed TUGC0 action to assure that such-problems will be precluded from occurring in the future. J-103
SEP 18 iga Mr. M. D. Spence This request is submitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding information/ evaluationFurther needs that could potentially affect the safe operation of their plant. requests for additional information of f.his nature will be made, if necessary, as the activities of the TRT progress. (Sincerely, *
,1 .-
~ :o . t; i.
Jt*'t)lIG;.Eisenhut.'DiYector
,6a'rre Division of Licensing, NRR
Enclosure:
As stated cc w/ enclosure See next page f J-104
COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Comission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels & Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallas, Texas 75201 B. R. Clements Mr. H. R. Rock Vice President Nuclear Gibbs and Hill, Inc. Texas Utilities Generating Company Skyway Tower 393 Seventh Avenue New York, New York 10001 400 North Olive Street L. B.'81 Dallas,' Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq. P. O. Box 355 1200 New Hampshire Avenue, N. W. Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas 78711 1901 Que Street, N. W. Mrs. Juanita Ellis, President ' Citizens Association for Sound David R. Pigott, Esq. Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq. CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W. San Francisco, California 94111 Suite 611 Washington, D. C. 20036 J-105
ENCLOSURE 1
- REQUEST FOR ADDITIONAL INFORMATION I. Electrical / Instrumentation Area
- a. Electrical Cable Teminations The Technical Review Team (TRT) inspected random samples of safety-related teminations, butt splices inside panels, and
- vendor-installed teminal lugs in General Electric (GE) motor control centers, and reviewed documentation relative to the installations.
- 1. The TRT found a lack of awareness on the part of quality control (QC) electrical inspectors to document in the inspection reports when the installation of the " nuclear heat-shrinkable cable insulation sleeves" was required to be witnessed.
Accordingly, TUEC shall clarify procedural requirements and provide additional inspector training with respect to the areas ) in which nuclear heat-shrinkable sleeves are required on splices ' and assure that such sleeves are installed where required.
- 2. The TRT found inspection reports that did not indicate that the required witnessing of splice installation was done. Examples l
are as follows: , IR ET-1-0005393 IR ET-1-0005396 IR ET-1-0005394 IR ET-1-0006776 IR ET-1-0005395 IR ET-1-0014790 Accordingly, TUEC will assure that all QC inspections requiring witnessing for butt splices have been perfomed and properly documented; and verify that all butt splices are properly identified on the appropriate drawings and are physically identified within the appropriate panels.
- 3. The TRT found a lack of splice qualification requirements and provisions in the installation procedures to verify the operability of those circuits for which splices were being used.
Accordingly, TUEC shall develop adequate installation / inspection i procedures to assure that the wiring splicing materials are ' qualified for the appropriate service conditions, and that splices are not located adjacent to each other.
- 4. Selected cable teminations were found that did not agree with their locations on drawings. Examples are as follows:
i i J-106
Panel CPI-ECPRCB-14, Cable E0139880 Panel CP1-ECPRTC-16, Cable E0110040 Panel CPI-ECPRTC-16, Cable E0118262 Panel CPI-ECPRTC-27, Cable EG104796 Panel CPX-ECPRCV-01, Cable EG021856 Panel CPI-ECPRCB-02, Cable NK139853 (nonsafety) Accordingly, TUEC shall reinspect all safety-related and associated terminations in the control room panels and in the termination cabinets in the cable spreading room to verify that their locations are accurately depicted on drawings. Should the results of this reinspection reveal an unacceptable level of nonconformance to drawings, the scope of this reinspection effort shall be expanded to include all safety-related and associated terminations at CPSES.
- 5. The TRT found cases where nonconformance reports (NCRs) concerning vendor-installed terminal lugs in GE motor control centers had been improperly closed. Examples are NCR Nos.
E-84-01066 through NCR E-84-01076, inclusive. Accordingly, TUEC shall reevaluate and redisposition all NCRs related to vendor-installed terminal lugs in GE motor control centers.
- b. Electrical Equipment Separation The TRT reviewed the separation criteria between separate cables, trays and conduits in the main control room and cable spreading room in Unit 1, and the compatibility of the alectrical erection specifications with regulatory requirements. The TRT reviewed documentation and inspected random samples of separation between safety-related cables, trays and conduits and between them and nonsafety-related cables, trays and conduits.
- 1. In numerous cases, safety-related cables within flexible conduits inside main control room panels did not meet minimum separation requirements. Examples are as follows:
Panel CP1-EC-PRCB-02 Panel CP1-EC-PRCB-07 Panel CPI-EC-PRCP-06 Panel CP1-EC-PRCB-08 Panel CPI-EC-PRCB-09 Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room for Unit 1, that
- contain redundant safety-related cables within conduits, or safety and non-safety related cables within conduits, and either correct each violation of the separation criteria, or J-107
demonstrate by analysis the acceptability of the conduit as a barrier for each case where the minimum separation is not met.
- 2. In several cases, separate safety and nonsafety-related cables and safety and nonsafety-related cables within flexible conduits inside main control room panels did not meet minimum separation requirements (Table 1 identifies examples of these cases). No evidence was found that justified the lack of separation.
Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room of Unit 1, and either correct each violation of the separation criteria concerning separate cables and cables within flexible conduits, or demonstrate by analysis the adequacy of the flexible conduit as a barrier.
- 3. The TRT found that the existing TUEC analysis substantiating the adequacy of the criteria for separation between conduits and cable trays had not been reviewed by the NRC staff.
Accordingly TUEC shall submit the analysis that substantiates the acceptability of the criteria stated in the electrical erection specifications governing the separation between independent conduits and cable trays.
- 4. The TRT found two minor violations of the separation criteria inside panels CPI-EC-PRCB-09 and CPI-EC-PRCB-03 concerning a barrier that had been removed and redundant field wiring not meeting minimum separation. The devices involved with the barrier were FI-2456A, PI-2453A, PI-2475A, and IT2450, associated with Train A; and FI-2457A, PI-2454A, PI-2476A, and 17-2451, The field wiring was associated with associated with Train B.
devices HS-5423 of Train B and HS-5574, nonsafety-related. Accordingly, TUEC shall correct two minor violations of the separation criteria inside panels CP1-EC-PRCB-09 and CP1-EC-PRCP-03 concerning a barrier that had been removed and redundant field wiring not meeting minimum separation. J-108
Table 1 Examples of Cases of Safety or Nonsafety-Related Cables In Contact With Other Safety-Related Cables Within Conduits in Control Room Panels
- 1. Control Panel CPI-EC-PRCB Containment Spray System Cable No. Train Related Instrument EG139373 B(green) Undetemined E0139010 A (orange) Undetemined
- 2. Control Panel CP1-EC-PRCB Reactor Control System Cable No. Train Related Instrument EG139383 B (green) Reactor manual trip switch E0139311 A (orange) Undetemined
- 3. Control Panel CP1-EC-PRCP Chemical & Volume Control System Cable No. Train Related Instrument EG139335 Wgreen) LCV-112C E0139301 A (orange) Undetemined
- 4. Control Panel CP1-EC-PRCB Auxiliary Feedwater Control System Cable No. Train Related Instrument E0139753 A (orange) FK-2453A E0139754 A orange) FK-2453B E0139756 B green) FK-2454A EG139288 B green) FK-24548 l
J-109
s
- c. Electrical Conduit Supports The TRT examined the nonsafety-related conduit support installation in selected seismic Category I areas of the plant. The support installation for non-safety related conduits less than or equal to 2 inches was inconsistent with seismic requirements and no evidence could be found that substantiated the adequacy of the installation for nonsafety-related conduit of any size. According to Regulatory Guide 1.29 and FSAR Section 3.7B.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the function of safety-related components or cause injury to plant personnel.
Accordingly. TUEC shall propose a program that assures the adequacy of the seismic support system installation for nonsafety-related conduit in all seismic Category I areas of the plant as follows:
- 1. Provide the resul.ts of seismic analysis which demonstrate that all nonsafety-related conduits and their support systems, satisfy the provisions.of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. Verify that nonsafety-related conduits less than or equal to 2
,' inches in diameter, not installed in accordance with the requirements of Regulatory Guide 1.29, satisfy applicable design requirements.
- d. Electrical QC Inspector Training / Qualifications ,
The TRT examined electrical QC inspector training and certification files, and requirements for personnel testing, on-the-job training, and racertification. The TRT also interviewed selected electrical QA/QC personnel.
- 1. 'The TRT found a lack of supportive documentation regarding personnel qualifications in the training and certification.
files, as required by procedures and regulatory requirements.. Also, the TRT found a lack of documentation for assuring that the requirements for electrical QC inspector recertification were being met. Specific examples are:
- One case of no documentation of a high school diploma or General Equivalency Diploma.
k e J-110 ,
\ 'g
- One case of no documentation to waive the remaining 2 months of the required 1 year experience.
- One case where a QC tecanician had not passed the required color vision examination administered by a professional eye specialist. A makeup test using colored pencils was administered by a QC supervisor, was passed, and then a waiver was given.
- Two cases where the experience requirements to become a Level 1 technician were only marginally met.
- One case of no documentation in the training and certification files substantiating that the person met the experience requirements.
Accordingly, TUEC shall review all the electrical QC inspector training, qualification, certification and recertification files against the project requirements and provide the information in such a form that each requirement is clearly shown to have been met by each inspector. If an inspector is found to not meet the training, qualification, certification, or recertification requirements, TUEC shall then review the records to determine the adequacy of inspections made by the unqualified individuals and provide a statement on the impact of the deficiencies noted on the safety of the project.
- 2. The TRT found a lack of guidelines and procedural requirements for the testing and certifying of electrical QC inspectors. Specifically, it was found that:
- No time limit or additional training requirements existed between a failed test and retest.
No controls existed to assure that the same test would not be given if an individual previously failed that test.
- No consistency existed in test scoring.
No guidelines or procedures were available to control the disqualification of questions from the test. No program was available for establishing new tests (except when procedures changed). The same tests had been utilized for the last 2 years. Accordingly, TUEC shall develop a testing program for electrical QC inspectors which provides adequate administrative guidelines, procedural requirements and test flexibility to assure that suitable proficiency is achieved and maintained. J-111 _ . . _ . - . _ . . . . _ , _ . _ ._ _ _ ~ , _ . _ _ . . .. _ . . _ . _ . _ . . _ _ _ _ _ _ . _ _ _ .
1 The deficiencies identified with the electrical QC inspections The have generic implications to other construction disciplines. implications of these findings will be further assessed as part of 4 the overall programatic review of QC inspector training and qualification and the results of this review will be reported under the QA/QC category on " Training and Qualification." II. Civil / Structural Area 1 a. Unable to Justify Reinforcing Steel Omitted in the Reactor Cavity The TRT investigated a documented occurrence in which reinforcing steel was omitted from a Unit I reactor cavity concrete placement This between the 812-foot and 819-foot i-inch elevations. reinforcement was installed and inspected according to drawing 2323-S1-0572, Revision 2. However, after the concrete was placed, Revision 3 to the drawing was issued showing a substantial increase in reinforcing steel over that which was installed. Gibbs & Hill Engineering was infomed of the omission Gibbs &by HillBrown & Root Engineering Nonconformance Report CP-77-6. replied that the omission in no way impaired the structural integrity of the structure. Nevertheless, the additional reinforcing steel was ' added as a precaution against cracking'which might occur in the vicinity of the neutron detector slots should a loss of coolant accident (LOCA) occur. A portion of the omitted reinforcing steel was also placed in the next concrete lift above the 819-foot t-inch level. This was done to partially compensate for the reinforcing steel omitted in the previous concrete lift and to minimize the overall area potentially subject to cracking. The TRT requested documentation indicating that an analysis was performed supporting the Gibbs & Hill conclusion. The TRT was subsequently infomed that an analysis had not been performed. Therefore, the TRT cannot determine the safety significance of this issue until an analysis is performed verifying the adequacy of the reinforcing steel as installed. Accordingly, TUEC shall provide an analysis of the as-built condition of the Unit I reactor cavity that verifies the adequacy o'f the i reinforcing steel between the 812-foot and 819-foot 1-inch elevations. The analysis shall consider all required load combinations. b., Falsification of Concrete Compression Strength Test Results The TRT investigated allegations that concrete strength tests were falsified. 50-445/79-09; The TRT reviewed an NRC Re 50-446/79-09)gion of this matter that IVincluded investigation (IE Report No. l J-112
interviews with fifteen individuals. Of these, only the alleger and one other individt.al stated they thought that falsification occurred, but they did not know when or by whom. The TRT also reviewed slump and air entrainment test results of concrete placed February during)the 1977 andperiod did notthe alleger find was employed any apparent variation(January in the 1976 to unifomity of the parameters for concrete placed during this period. Although the uniformity of the concrete placed appears to minimize the likelihood that low concrete strengths were obtained, other allegations were raised concerning the falsification of records associated with slump and air content tests. The Region IV staff addressed these allegations by assuming that concrete strength test results were adequate. Furthemore, a number of other allegations dealing with concrete placement problems (such as deficient aggregate grading and concrete in the mixer too long) were also resolved by assuming that concrete strength test results were adequate. The TRT agrees with Region IV that, while the preponderance of evidence - suggests that falsification of results did not take place, the matter cannot be resolved completely on the basis of concrete strength test results, especially if there is any doubt about whether they may have been falsified. Due to the importance of the concrete strength test results, the TRT believes that additional action by TUEC is necessary to provide confimatory evidence that the reported concrete strength test results are indeed representative of the strength of the concrete installed in the Category I concrete structures. Accordingly, TUEC shall detemine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a program to assure acceptable concrete strength. The program shall include tests such as the use of random Schmidt hamer tests on the concrete in areas where safety is critical. The program shall include a comparison of the results with the results of tests per-formed on concrete of the same design strength in areas where the strength of the concrete is not questioned, to determine if any significant variance in strength occurs. TUEC shall submit the program for perfoming these tests to the NRC for review and approval prior to perfoming the tests.
- c. Maintenance of Air Gap Between Concrete Structures The TRT investigated the requirements to maintain an air gap between concrete structures. Based on the review of available inspection reports and related documents, on field observations, and on discussions with TUEC engineers, the TRT cannot detemine whether an adequate air gap has been provided between concrete structures. Field investigations by B&R QC inspectors indicated
,i unsatisfactory conditions due to the presence of debrf,s in the air I l J-113 l
gap, such as wood wedges, rocks, clumps of concrete and rotofoam. The disposition of the NCR relating to this matter states that the
" field investigation reveals that most of the material has been removed." However, the TRT cannot determine from this report (NCR C-83-01067) the extent and location of the debris remaining between the structures.
Based on discussions with TUEC engineers, it is the TRT's understanding that field investigationsInwere addition, made itbut is that no not apparent permanent records were maintained. that the permanent installation of elastic joint filler material ("rotofoam") between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the seismic analysis assumptions and dynamic models used to analyze the buildings, as these analyses are delineated in the Final Safety Analysis Report (FSAR). The TRT, therefore, concludes that TUEC has not adequately demonstrated compliance with FSAR Sections 3.4.1.1.1, 3.8.4.5.1, and 3.7.8.2.8, which require separation of Seismic Category I buildings to prevent seismic interaction during an earthquake. Accordingly TUEC shall:
- 1. Perform an inspection of the as-built condition to confinn that adequate separation for all seismic category I structures has been provided.
- 2. Provide the results of analyses which demonstrate that the presence of rotofoam and other debris between all concrete structures (as detennined by inspections of the as-built conditions) does not result in any significant increase in seismic response or alter the dynamic response characteristics of the Category I structures, components and piping when compared with the results of the original analyses.
- d. Seismic Design of Control Room Ceiling Elements The TRT investigated the seismic design of the ceiling elements installed in the control room. The following matrix designates those ceiling elements present in the control room and their seismic category designation:
J-114 7 r
- 1. Heating, Ventilating and Air - Seismic Category I Conditioning - Seismic Category I
- 2. Safety-Related Conduits - Seismic Category II
- 3. Nonsafety-Related Conduits Lighting Fixtures - Seismic Category II
- 4. - Non-Seismic
- 5. Sloping Suspended Drywall Ceiling
- 6. Acoustical Suspended Ceiling - Non-Seismic
- 7. Lowered Suspended Ceiling - Non-Seismic According to Regulatory Guide 1.29 and FSAR Section 3.78.2.8, the seismic Category II and nonseismic items should be designed in such a 4
way that their failure would not adversely affect the functions of safety-related components or cause injury to operators. For the nonseismic items (other than the sloping suspended drywall ceiling),andfornonsafety-relatedconduitswhose diameter is 2 inches or less, the TRT could find no evidence that the possible effects of a failure of these items had been : considered. In addition, the TRT detennined that calculations for ' i seismic Category II components (e.g., lighting fixtures) and the calculations for the sloping suspended drywall ceiling did not : l adequately reflect the rotational interaction with the nonseismic i items, nor were the fundamental frequencies of the supported ' masses detennined to assess the influence of the seismic response spectrum at the control room ceiling elevation would have on the seismic response of the ceiling elements. Accordingly, TUEC shall provide:
- 1. The results of seismic analysis which demonstrate that the 1
nonseismic items in the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. An evaluation of seismic design adequacy of support i l
systems for the lighting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) l which accounts for pertinent floor response characteristics of the systems.
- 3. Verification that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
- 4. The results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
J-115
- 5. The results of an analysis which demonstrate that the foregoing problems are not applicable to other Category II and nonseismic structures, systems and components elsewhere in the plant.
- e. Unauthorized Cutting of Rebar in the Fuel Handling Building The TRT investigated an alleged instance of unauthorized cutting of rebar associated with the installation of the trolley process aisle rails in the Fuel Handling Building. The claim is that during installation of 22 metal plates in January 1983, a core drill The TRT was used to drill abcut 10 holes approximately 9 inches deep.
reviewed the reinforcement drawings for the Fuel Handling Buf1 ding and determined that there were three layers of reinforcing steel in the top reinforcement layer of the slab. This reinforcement layer consisted of a No.18 bar running in the east-west direction in the first and third layers, and a No. 11 bar running in the north-south direction on the second layer. The review also revealed that the layout of the reinforcement and the trolley rails was such that the east-west reinforcement would interfere with the drilling of holes along only one rail location. However, if 9-inch holes were drilled, both the first and third layers of No. 18 reinforcement would be cut. Design Change Authorization No. 7041 was written for authorization to cut the uppermost No.18 bar at only one rail location, but did not reference authorization to cut the lower No. 18 bar. DCA-7041 also stated that the expansion bolts and base plates may be moved in the east-west direction to avoid interference with reinforcement running in the north-south direction. The information, described in DCA-7041, was substantiated by Gibbs & Hill calculations. If the ten holes were actually drilled 9 inches deep, then the allegation that the reinforcement was cut without propar authorization would be valid. Accordingly, TUEC shall provide:
- 1. Information to demonstrate that only the No. 18 reinforcing steel in the first layer was cut, or
- 2. Design calculations to demonstrate that structural integrity is maintained if the No. 18 reinforcing steel on both the first and third layers was cut.
III. Test Programs Area
- a. Hot Functional Testing (HFT)
The TRT reviewed a sample of the completed data packages for HFT preoperational test procedures, pertinent startup adninistrative procedures, NRC inspection reports, and the preoperational test index and its schedule. The TRT also inspected test deficiency reports J-116 _ . . _ _ - _ _ _ - - _ - - . -_ . ~__.. _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ . ~ . _ _ _ _ . _ , _ _
-B l
(TDRs) that were generated as a result of test deficiencies found prior to and during HFT. 1. Chapter 14 of the FSAR and Regulatory Guide 1.68 provide requirements for the conduct of preoperation test objectives were not met. Group approved incomplete data packages for at least three 4 preoperational hot functinal tests. These were: Deficiency Test Procedure ) I Because acceptable voltages ICP-PT-02-12. " Bus Voltage and Load Survey" could not be achieved with changed. A subsequent engineering evaluation required returning to the original taps, but no retest was perfomed. Level detectors 1-LT-517, 518 ICP-PT-34-05, " Steam and 529 were rep, laced with Generator Narrow Range temporary equipment of a . Level Verification" design that was different from that which was to be eventually installed Level detector 1-LT-461 appeared 1CP-PT-55-05 to be out of calibration during the
" Pressurizer Level test and was replaced after the test.
Control" The retest approved by the JTG was a cold calibration rather than a test consistent with the original test objective, which was to obtain satisfactory data under hot conditions. Accordingly, TUEC shall review all comp objectivesThe were not met, or prerequisite conditions were three items identified by the TRT shall be satisfied. included, along with appropriate justification, in the test deferral packages presented to the NRC. i J-117
l
- 2. The TRT noted during a review of HFT completed test data that the JTG did not approve the data until after cooldown from the test. The tests are not considered complete until this approval is obtained. In order to complete the proposed post-fueling, deferred preoperational HFT, the JTG, or a similarly qualified group, must approve the data prior to proceeding to initial criticality. The TRT did not find any document providing assurance that TUEC is connitted to do this.
Accordingly, TUEC shall connit to having a JTG, or sim'larly qualified group, review and approve all post-fueling preoperational test results prior to declaring the system operable in accordance with the technical specifications.
- 3. The TRT pointed out that in order to conduct preoperational tests at the necessary temperatures and pressures after fuel load, certain limiting conditions of the proposed technical specifications cannot be met, e.g., all snubbers will not be operable since some will not have been tested.
Accordingly, TUEC shall evaluate the required plant conditions for the deferred preoperational tests against limiting conditions in the p'roposed technical specifications and obtain NRC approval where deviations from the technical specifications are necessary.
- 4. Data for the thermal expansion tests (which have not yet been approved by the JTG) did not provide for traceability between the calibration of the measuring instruments and the monitored locations, as required by Startup Administrative Procedure-7.
The information was separately available in a personal log held by Engineering. Accordingly, TUEC shall incorporate the information necessary to provide traceability between thermal expansion test monitoring I locations and measuring instruments. TUEC shall also establish administrative controls to assure appropriate test and me w .ing equipment traceability during future testing. l 1
- b. Containment Intergrated Leak Rate Testing (CILRT)
The TRT reviewed the data package for the CILRT perfonned on Unit 1, and discussed the conduct of the test with TUEC and NRC personnel who participated in or witnessed it. l J-118
i It was Apparently after repairing leaks found during the firsttwo atte successfully completed after three electrical penetrations were isolated because the leakage through them could not be stopped. Though the leaks were subsequently Inrepaired and individ 1 i to perfom the CILRT with these penetrations isolated. addit ! which is neither endorsed by the NRC nor in accordance with FSAR !
)
comitments. l Accordingly, TUEC shall identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rath ANSI N45.4-1972. deviations from FSAR consnitments.
- c. Prerequisite Testing The TRT reviewed FSAR commitments, startup administrative procedures, ,
prerequisite test records, craft personnel qualification records, The TRT and ' discussed them with startup and craft management personnel. also observed test support craft personnel at work and interviewed seine of them to gain familiarity with their attitudes and capabilities. j The review of test records revealed that craft personnel were signing to verify initial conditions for tests in" Conduct violationof of startup Testing" Administrative Procedure-21, entitled:This procedure requires this func (CP-SAP-21). ' by System Test Engineers (STE). Startup management had issued a memorandum improperly authorizing craft personnel to perfom these
- verifications on selected tests.
Accordingly, TUEC shall rescind the startup memorandum (STM-83084), f which was issued in conflict with CP-SAP-21, and ensure that no other l memoranda were issued which are in conflict with approved procedures.
- d. Preoperational Testing l
The TRT assessed the preoperational test program by reviewing l administrative procedures, interviewing startup personnel, and i examining test records, schedules, system assignments, subsystem i definition packages, and the master data base. ProblemsThe found TRT with test that also found data arewere STEs addressed in section not being provided with III.a of this enclosure. j current design infomation on a routine, controlled basis, and had to update their own material when they considered it appropriate. l Accordingly. TUEC shall establish measures to provide gre current controlled design documents and change notices. r J-119 l -- ---_- _-_ - - - - - - - . _ . , _ _ . ,, _
= - - _
Attachment 4 Occket Nos.: 50-445 and 50-446 00T 5 1!$4 Texas Utilities Electric Company
; Attn: M. D. Spence, President, TUGC0 J
Skyway Tower 400 North Olive Street
- Lock Box 81 l
Dallas, Texas 75201 1
Dear Mr. Spence:
Subject:
September llr, 1984 Letter, D. G. Eisenhut to M. D. Spence, Re: Comanche Peak Review During our meeting on September"18, 1984 at Bethesda, Maryland, we discussed the technical issues regarding Comanche Peak which the NRC Technical Review Team identified as having potential safety implications and thus requiring additional .information. The subject letter listing these items and the information that we requested were provided to you during that meeting. + We have since discovered some typographical errors in the Enclosure to the September 18, 1984 letter and provided Mr. John Merritt of your staff with a marked-up copy of that letter on September 21, 1984. Enclosed for your information is an errata to the letter. Sincerely. Original sipeW j Darrell G. EiseM ! Darrell G. Eisenhut, Director Division of Licensing ' Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure: See next page 4
]
i l J.120
--c--. +, - . . - - _ --., -~.,.-,.-4,y, , + - - - - + . , - - . , - , - - - .
, - . s.-+-c-,--, -,_, ,,,-w .+ >-m , m. ,- 4,r,.--e-. .%-m- ---.--r----m - - --
Enclosure Errata To Enclosure 1 to September 18, 1984 Letter, D. G. Eisenhut to M. D. Spence
- 1. Page 2, line 1
! Panel cpl-ECPRCB-la should be Panel cpl-ECPRCB-04
- 2. Page 2, 8th line from bottom of page 4
Panel cpl-EC-PRCP-06 should be Panel cpl-EC-PRCB-06
- 3. Page 4, item 3 Control Panel cpl-EC-PRCP-06 should be Control Panel cpl-EC-PRCB-06
- 4. Page 9, 3rd line from bottom of first full paragraph
! Sections 3.4.1.1.1 i should be 4
Sections 3.8.1.1.1
- 5. Page 10, top of page, item 7 Lowered Suspended Ceiling should be Louvered Suspended Ceiling l
i I 4 I 4 J-121
,----,--,,re,..,,-- ,-.--,..-,,.,-,n. -
CMANCHE PEAX . Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector /Ccmanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. 5. Nuclear Regulatory Washington, D. C. 20036 Commission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels & Wooldridge Mr. John T. Collins 2001 Bryan Tcwer, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Mcmer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallas, Texas 75201 B. R. Clements . Vice President Nuclear Texas Utilities Generating Ccmpany Mr. H. R. Rock Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker William A. Burchette, Esq. Westinghouse Electric Corporation 1200 New Hampshire Avenue, N. W. P. 0. Box 355 - Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. . Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas 78711 1901 Que Street, N. W w
. Washington, D. C. 20)09 f Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Mont9cmery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roiiman, Esq. CYGNA Trial Lawyers for Public Justice 101 Califcenia Street 2000 P. Street, N. W. San Francisco, California 94111 Suite 611 Washington, D. C. 20036 J-122
7,C,,,Omu un v i NvCa A uGutATo-v Co u 55 oh i u ogNvust,.u co.,r,oC.-,m u..,..w NUREG-0797 BIBLIOGRAPHIC DATA SHEET Supplement No. 7 2L ... TaTLE AND SveisTLt 4 RECi'iENT 5 ACCES$ SON NUM8ER iafety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 '^"'""'" January I,,,,1985 7 DATE REPORT ISSUED MONTH January IVEAR 1985 9 PRQJECT T A5suwCRn UNIT NUM0ER
*E A80AMING ORG ANa2 ATION NAME AND M AILING ADORE 55 flewee le Co887
- livision of Licensing
)ffice of Nuclear Reactor Regulation 'o "* *v***a 1.5. Nuclear Regulatory Commission fashington, DC 20555 5FON50*NG ORG ANi2 ATION N AME AND W AILING ADDRE SS (lac,ver le Cod.1 12a TYPE OF REPORT TECHNICAL tame as 8. above i2. PEmiOO COvEno ,,,.c . ,
July 1984 - January 1985
$UPPLEMENTARY NOYt3 locket Nos. 50-445 and 50-446 Ansta ACT L200 were. er ,e.no lupplement No. 7 to the Safety Evaluation Report for the Texas Utilities Generating
'ompany application for a license to ope. rate the Comanche Peak Steam Electric Station ocated in Somervell County, Texas has been jointly prepared by the Office of Nuclear 'eactor Regulation and the Technical Review Team of the U. S. Nuclear Regulatory Com-iission. This Supplement provides the results of the staff's evaluation and resolution 'f approximately 80 technical concerns and allegations in the areas of Electrical / nstrumentation and Test Program regarding construction practices at the Comanche 'eak facility. i l KEY WOA05 AND DOCuwf NT ANALvist
- 16. DESCRePTOR$
AvasLAsstiTV $7 ATEwf NT 17 SECURITY CLA51so sCAfeON 18 NvuSER 08PAGES UfkU5$1FIED 19 SECURITY CL A881FsCATION 20 PRICS NLIMITED ONCCA$SIFIED s
NUREG-0797 Supplement No.8 Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446 Texas Utilities Generating Company, et al. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation February 1985 p" "%,, Rws.e
- s. .
ys w g a rd !:0lA-85-59 I 6 124
t ABSTRACT Supplement 8 to the Safety Evaluation Report for the Texas Utilities Electric Company application for a license to operate Comanche Peak Steam Electric Sta-tion, Units 1 and 2 (Docket Nos. 50-445, 50-446), located in Somervell County, Texas, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Comanche Peak Technical Review Team of the U. S. Nuclear Regulatory Commission. This Supplement provides the results of the staff's evaluation and resolution of approximately 80 technical concerns and allegations relating to civil and structural and miscellaneous issues regarding construction and plant readiness testing practices at the Comanche Peak facility. Issues raised during recent Atomic Safety and Licensing Board hearings will be dealt with in i future supplements to the Safety Evaluation Report. W l
- (
( i i
. l 4
I Comanche Peak SSER 8 fii I i
S 4
,i 4
TABLE OF C0'NTENTS Page
- ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii l
ACRONYMS AND ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . vii i
- 1. INTRODUCTION. ........... .............. 1-1 The Comanche Peak Technical Review Team for SER Supplement 8. . 1-2
! APPENDIX K - Status of Staff Evaluation and Resolution of Technical Concerns and Allegations Relating to Civil and Structural and Miscellaneous Issues Regarding Construction and Plant Readiness Testing Practices at Comanche Peak Steam Electric Station,
- Units 1 and 2 ... . . . . . . . . . . . . . . . . . . . K-1 i I t i i i i I Comanche Peak SSER 8 v
l l ACRONYMS AND ABBREVIATIONS AA - independent assessment program allegation AB - American Bridge AB - bolt allegation ,- ABRR - as-built reverification records A-C - Allis-Chalmers AC - concrete /rebar allegation ACI - American Concrete Institute AD - design of pipe / pipe support allegation - ADS - audit discrepancy report AE - electrical allegation AE00 - Office for Analysis and Evaluation of Operational Data (NRC) AFW - auxiliary feedwater system AH - hanger allegation AI - intimidation allegation AISC - American Institute of Steel Construction ALARA- as low as reasonably achievable AM - miscellaneous allegation ANI - authorized nuclear inspector ANS - American Nuclear Society ANSI - American National Standards Institute AO - protective coating allegation AP - pipe and pipe support allegation APC - AMP Product Corporation AQ quality assurance / quality control allegation AQB QA/QC bolt allegation AQC QA/QC concrete /rebar allegation AQE QA/QC electrical allegation AQH QA/QC hanger allegation AQL acceptable quality level AQO QA/QC coating allegation AQP QA/QC pipe and pipe support allegation AQW QA/QC welding allegation ARMS - Automated Records Management System ASLB - Atomic Safety and Licensing Board ASME - American Society of Mechanical Engineers ASTM - American Society for Testing and Materials AT - acceptance test AT - test program allegation AV - vendor / generic allegation AW - welding allegation B&PVC - Boiler & Pressure Vessel Code B&R - Brown & Root, Inc. BNL - Brookhaven National Laboratory BRHL - Brown & Root Hanger Locations BRIR - Brown & Root Inspection Report Comanche Peak SSER 8 vii
BRP - Brown & Root piping isometric drawing BTP - Backfit Test Program BWR - boiling water reactor C&L - Corner and Lada (computer program) C&S - civil and structural CAR - Corrective Action Request CASE - Citizens Association for Sound Energy CAT - Construction Appraisal Team (NRC) CB&I - Chicago Bridge & Iron Company CCL - Corporate Consulting and Development Company, Limited CCS - Component Cooling System CCW - component cooling water CEL - Coating Exempt Log CFR - Code of Federal Regulations CHN - construction hold notice CILRT - containment integrated leak rate test CMC - component modification cards CMTR - certified material test report COT - construction operation traveler CP - Comanche Peak CP - construction permit CPPE - Comanche Peak Project Engineering CPSES - Comanche Peak Steam Electric Station CPSIG - Comanche Peak Seismic Interaction Group CSTS - Construction and Startup/ Turnover Surveillance Group (TUEC) CVCS - chemical and volume control system CZ Carboline Carbo zine 11 DBA - design basis accident DCA - design change authorization DCC - Document Control Center (TVEC) DCTG - Design Change Tracking Group DCVG - Design Change Verification Group DE - Division of Engineering (NRC) DFT - dry film thickness DL - Division of Licensing (NRC) 0-6 - Ameron Dimetcote 6 E&I - Electrical and Instrumentation ECCS - emergency core cooling system . EDO - Executive Director for Operations (NRC) ERG - emergency response guideline ETG - Electrical Test Group (TUEC) FDSG - Field Damage Study Group (TVEC) FJO - field job orders FP - fire protection FSAR - Final Safety Analysis Report FW - field weld Comanche Peak SSER 8 viii
G&H - Gibbs & Hill GAP - Government Accountability Project GDC - general design criteria GE - General Electric Corporation GED - General Equivalency Diploma GHH - Gibbs & Hill hanger (isometric drawing) HFT - hot functional test HIR - hanger inspection report HP - hanger package HP - high pressure HVAC - heating, ventilation and air conditioning system HX - heat exchangers IAP - Independent Assessment Program ICC - inadequate core cooling IE - Office of Inspection and Enforcement (NRC) IEB - Inspection and Enforcement Bulletin IEEE - Institute of Electrical and Electronics Engineers IM - INP0 - interoffice memorandum (TUEC) Inctitute for Nuclear Power Operations IOM - interoffice memorandum IR - inspection report (NRC) IRN - item removal notice ITT-G - ITT Grinnell JTG - Joint Test Group (TUEC) JUHA - Joint Utility Management Assessment Group LE - left end LOCA - loss of coolant accident LP - liquid penetrant M&P - mechanical and piping MAR - maintenance action request MCC - MDB - motor control center (GE) master data base MIFI - mechanical fabrication inspector MIL - material identification list (or log) MIME - Mechanical Equipment Inspector HQE Mechanical Quality Engineering MR - material requisition MRS manufacturer's record sheet MS - main steam (line) MWDC - multiple weld data card N/A - not applicable NCR nonconformance report (TUEC) Comanche Peak SSER 8 ix
NDE - nondestructive examination NOT - nondestructive testing NI - never incorporated NONSAT - nonsatisfactory NOV - Notice of Violation (NRC)
' NPSH - net positive suction head NPSI - Nuclear Power Service Incorporated NRC - U.S. Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation (NRC)
NSSS - noclear steam supply system a O&M - Operations and Maintenance (TVEC) OBE - operating basis earthquake 01 - Office of Investigations (NRC) OJT - on-the-job training OL operating license ORNL - Oak Ridge National Laboratory PC - protective coating PCR - plant change request PET
- permanent equipment transfer PFG
- Paper Flow Group PFS
- pipe fabrication shop PORV - power operated relief valve PPM - parts per million PSAR - Preliminary Safety Analysis Report PSE
- Pipe Support Engineering (TUEC)
PT
- preoperational test PTS
- pressurized thermal shock PWR - pipe whip restraints PWR - pressurized water reactor i P-305 - Carboline Phenoline 305 QA quality assurance QAI
- quality assurance investigation (TVEC)
QC quality control QE quality engineer RCB
- Reactor Containment Building RE -
right end RES
- Office of Nuclear Regulatory Research (NRC)
RFIC - request for information or clarification (B&R) RG
- Regulatory Guide (NRC)
RHRS - residual heat removal system RI - NRC Region I Office RIR - receipt inspection report (TUEC) RIV - NRC Region IV Office RPE
- radiation protection engineer RPI - rod position indication RPS
- radiation protection supervisor i l x
Comanche Peak SSER 8
l I RPS - report process sheet (TUGCO) RPV - reactor pressure vessel RPVRI - reactor pressure vessel reflective insulation RRI - Resident Reactor Inspector (NRC) RV - reactor vessel RWN - room work notifications SAP - startup administration procedure SALP - Systematic Assessment of Licensee Performance (NRC) SAT - satisfactory SAVC - structural assembly verification card SER Safety Evaluation Report (NRC) SI - safety injection SIS - Special Inspection Services SMAW - shielded metal arc welding SNM - special nuclear material SORC - Station Operations Review Committee SRIC - Senior Resident Inspector for Construction (NRC) SRP - SRT - Standard Review Plan (NRC) Special Review Team (NRC) SSE safe shutdown earthquake SSER - Safety Evaluation Report Supplement SSI - safe shutdown impoundment SSPC - Steel Structures Painting Council SSWP - station service water pumps STE - system test engineer SWA - startup work authorization SWO - shop work order TDCR - test deficiency change request TDI - Transamerica Delaval, Inc. TDR - test deficiency report 10 CFR 50 - Title 10 Code of Federal Regulations Part 50 TI - temporary instruction TIDC - THE - Division TUEC of Technical Nuclear EngineeringInformation and Document Control (NRC) TP - test program
. TPD -
test procedure deviation Tr - transcript TRT - Technical Review Team (NRC) TSABC - TSDR - technical services as-built coordinator technical services design review coordinator TSI - thermolag TSMO - Technical Services Mechanical Drafting TSP - tri sodium phosphate TUEC - Texas Utilities Electric Company TUGC0 - Texas Utilities Generating Company TUSI - Texas Utilities Service, Inc. UCC - University Computing Company 4 USI - unresolved safety issue i UT - ultrasonic test UTA - University of Texas at Austin i Comanche Peak SSER 8 xi
t h VCD - vendor-certified drawing VT - visual weld (inspector) W - Westinghouse Electric Corporation
~C weld data card 4
WO WFML - weld filler metal log WPS - welding procedure specification l i i i i i i i 5 1 i I l t i t I 1 l i 1 1 I i i 1 1
)
i f i f f 5 L I [ , i Comanche Peak SSER 8 xii i s '-,..,__.-.-._.._,...--.-.-..m.. - . - - - . - . . . - _ _ _ . _ , . . - _ . - - - _ - - - _ . . . , . _ . . _ . , . _ , . . _ _ - - . . . _._....._...__!
l l 1 INTRODUCTION On July 14, 1981, the U.S. Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) (NUREG-0797) related to the application by the Texas Utilities Electric Company (TVEC) for a license to operate Comanche Peak Steam Electric Station (CPSES) Units 1 and 2. Subsequently.seven supplemental Safety Evaluat' ion Reports (SSERs) were issued by the staff. Supplement No. 7, pub-lished in January 1985, dealt with technical concerns and allegations in the electrical and instrumentation and test program areas about Comanche Peak. This report, Supplement No. 8, is the second of a series of SSERs dealing with various technical concerns and allegations about Comanche Peak. This report addresses approximately 80 technical concerns and allegations relating to civil and structural and miscellaneous issues. Appendix K to this report provides details of the staff's evaluation and findings of'these technical concerns and allegations. The technical concerns and allegations about Comanche Peak were part of the regulatory issues that remained outstanding toward the completion of construc-tion of the Comanche Peak facility. The NRC's Executive Director for Opera-tions (EDO) issued a directive on March 12, 1984, establishing a program for assuring the overall coordination / integration of these issues and their reso-lution prior to the staff's licensing decision. In response to the E00's directive, a program plan was developed and approved on June 5, 1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of NRC's Region IV Office. This pro-gram plan, entitled Comanche Peak Plan for the Completion of Outstanding Regu-latory Actions, specified the critical path issues, addressed the scope of work needed, and provided a project schedule for completion. Management and coordination of all the outstanding regulatory actions for Comanche Peak are under the overall direction of Mr. Vincent S. Noonan, the NRC Comanche Peak Project Director. Mr. Noonan may be contacted by calling 301-492-7903 or by writing to the following address: 3 Mr. Vincent S. Noonan Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555 5 Copies of this Supplement are available for public inspection at the NRC's Public Document Room at 1717 H Street, NW, Washington, D.C. 20555, and the i Local Public Document Room, located at the Somervell County Public Library On the Square, P.O. Box 1417, Glen Rose, Texas, 76043. Availability of all material cited is described on the inside front cover of this report. l Comanche Peak SSER 8 1-1
l The Comanche Peak Technical Review Team for SER Supplement 8 Bangart, R. - Region IV, NRC I Brown, C. - EG&G, San Ramon ( EG&G, Idaho Falls J Corbett, J. - Devers, J. - Parameter Ellershaw, L. - Region IV, NRC Gagliardo, J. - Reactor Training Center, IE, NRC Haughney, C. - COMEX Corporation Hofmayer, C. - Brookhaven National Laboratory Hunnicutt, D. - Region IV, NRC Ippolito, T. - Office of Analysis and Evaluation of Operational Data, NRC Jeng, D. - Office of Nuclear Reactor Regulation, NRC j Jones, L. - EG&G, Idaho Falls Kelly, J. - Region IV, NRC Langowski, T. - ETEC i Noonan, V. - Office of Nuclear Reactor Regulation, NRC I Oliu, W. - Division of Technical Information and Document Control, NRC Payne, B. - EG&G, Idaho Falls ! Philleo, R. - Parameter Phillips, S. Region IV, NRC Saffell, B. - Battelle Columbus Laboratories Shao, L. - Office of Nuclear Regulatory Research, NRC Skow, M. - Region IV, NRC Smith, W. - Region IV, NRC Tang, R. C. - Office of Nuclear Reactor Regulation, NRC Tapia, J. - Region IV, NRC Vietti, A.
- Office of Nuclear Reactor Regulation, NRC Warren, F. - EG&G, Idaho Falls Wessman, R. - Office of Nuclear Reactor Regulation, NRC Wise, R. - Region IV, NRC Zudans, J. - Office of Inspection and Enforcement, NRC 1-2 Comanche Peak SSER 8
l APPENDIX K STATUS OF STAFF EVALUATION AND RESOLUTION OF TECHNICAL CONCERNS AND ALLEGATIONS RELATING TO CIVIL AND STRUCTURAL AND MISCELLANEOUS ISSUES REGARDING CONSTRUCTION AND PLANT READINESS TESTING AT COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2
TABLE OF CONTENTS PaSe
- 1. Introduction..................................................... K-1
- 2. Comanche Peak Technical Concerns and Allegations Management Program........................................................ K-3 2.1 Background.................................................. K-3
- 2. 2 Review Approach and Methodology............................. K-3 2.2.1 Concern and Allegation Tracking System............... K-3 2.2.2 Review Methodology................................... K-4 2.2.3 Interviews with Allegers............................. K-5 2.3 Communications with TUEC.................................... K-5
- 3. Summary of Evaluations........................................... K-7 3.1 Civil and Structural Group Summary.......................... K-7 3.1.1 Scope of Concerns and Allegations.................... K-7 3.1.2 Civil and Structural Group........................... K-9 3.1.3 Findings for Civil and Structural Issues............. K-10 3.1.4 Overall Assessment and Conclusions................... K-11 3.2 Miscellaneous Group Summary................................. K-11 3.2.1 Scope of Concerns and Allegations.................... K-11 3.2.2 Miscellaneous Group.................................. K-13 3.2.3 Findings for Miscellaneous Issues.................... K-13 3.2.4 Overall Assessment and Conclusions................... K-15
- 4. Actions Required of TUEC......................................... K-16 4.1 Civil and Structural Area................................... K-16 4.1.1 Rebar Improperly Installed or 0mitted. . . . . . . . . . . . . . . . K-16 4.1.2 Falsification of Concrete Compression Strength Test Results................................ K-16 4.1.3 Maintenance of Air Gap Between Concrete Structures........................................... K-16 4.1.4 Seismic Design of Control Room Ceiling Elements...... K-17 4.1.5 Unauthorized Cutting of Rebar In Fuel Heudling Building.................................... K-17 4.1.6 Hollow Places in Concrete Behind Unit 2 Reactor Cavity Liner................................. K-17 Comanche Peak SSER 8 K-iii
TABLE OF CONTENTS (continued) P.ag K-17 4.2 Miscellaneous Area.......................................... 4.2.1 Gap Between Reactor Pressure Vessel Reflective Insulation (RPVRI) and the Biological Shield Wall.... K-17 4.2.2 Control of Debris in Critical Spaces Between K-18 Components and/or Structures......................... K-18 4.2.3 Polar Crane Shimming................................. Attachments K-19 1 - Listing of Technical Concerns and Allegations in the Civil and Structural and Miscellaneous Areas................................ 2 - Assessment of Individual Technical Concerns and Allegations in K-27 Civil and Structural and Miscellaneous Areas...................... 3 - September 18, 1984, letter with enclosure, D. G. Eisenhut, Director, Division of Licensing, Office of Nuclear Reactor Regulation, NRC, to M. D. Spence, President, Texas Utilities K-149 Electric Company,
Subject:
Comanche Peak Review.................. 4 - October 5, 1984, letter with enclosure, D.G. Eisenhut, Director Division of Licensing, Office of Nuclear Reactor Regulation, NRC, to M. D. Spence, President, Texas Utilities Electric Company, subject: errata sheet for September 18, 1984, letter..... K-167 5 - November 29, 1984, letter with enclosure, D. G. Eisenhut, Direc-tor, Division of Licensing, Office of Nuclear Reactor Regulation, NRC, to M. D. Spence, President, Texas Utilities Electric Company, K-171
Subject:
Comanche Peak Review.................................... Comanche Peak SSER 6 K-iv
m
- 1. Introduction As construction of the Comanche Peak Steam Electric Station was nearing comple-tion, issues that remained to be resolved prior to the consideration of issuance of an operating license were complex, resource intensive, and spanned more than one NRC office. To ensure the overall coordination and integration of these issues, and to ensure their resolution prior to licensing decisions, the.NRC's Executive Director for Operations (EDO) issued a memoraadum on March 12, 1984, directing the NRC's Office of Nuc1 car Reactor Regulation to rr.anage all necessary NRC actions leading to prompt licensing decisions,'and assigning the Director, NRC's Division of Licensing, the lead responsibility for coordinating and inte-grating the related efforts of various offices within the NRC.
The principal areas needing resolution before a licensing decision on Comanche Peak can be reached include: (1) the completion and documentation of the staff's review of the Final Safety Analysis Report (FSAR); (2) those issues in contention before the NRC's Atomic Safety and Licensing Board (ASLB); (I) the completion of necessary NRC regional inspection actions; and (4) the completion and documentation of the staff's review of technical cco: erns and allegations regarding design and construction of the plant. Technical concerns and allegations about Comanche Peak, totalling approximately 900, have been raised mainly by the quality assurance / quality control (QA/QC) . personnel working or having worked on site. Their job responsibilities involve or involved QA/QC aspects of safety-related structures, systems, and components to determine whether and to what extent such items are manufactured, purchased, stored, maintaincd, installed, tested, and inspected as required by project documents and procedures. Many of these allegations were made orally to NRC Region IV staff, NRC Comanche Peak Site Resident Inspectors, NRC investigators, or in letters to the NRC, as well as in testimony before the Atomic Safety and Licensing Board (ASLB). Individuals with allcgations were also sponsored by the intervenor group Citizens Association for Sound Energy (CASE) and the Government Accountability Project (GAP). General allegations about poor con-struction work at Comanche Peak were also made in several newspaper articles in the Dallas / Fort Worth, Texas areas. By the end of April 1984, the staff identified approximately 400 technical con-cerns and allegations related to the construction of the Comanche Peak facility, including findings by NRC's Special Review Team. (See Section 2.1 below.) During its investigation of a concern or allegation, the TRT identified addi- > tional concerns. Interviews with allegers also yielded addi,t'ional concerns. By December 1984, approximately 600 concerns and allegations had been identified. In addition, approximately 300 allegations were recently provided to the TRT by one alleger. These technical concerns and allegations were' grouped by subject into the following areas: , Electrical and Instrumentation . Civil and Structural , 1 Comanche Peak SSER 8 K-1 I i a \
l
- Mechanical and Piping
- Quality Assurance and Quality control (QA/QC) ,
- Coatings
- Test Program Miscellaneous This report is the second of a series of reports dealing exclusively with the NRC staff's efforts to evaluate and resolve the technical concerns and allega-tions raised by various parties and individuals regarding construction practices at the Comanche Peak facility. Included in this report are civil and structural and miscellaneous issues. An allegation or concern was assessed as having no safety significance if, based on technical findings, the assessment showed that a structure, component, or system would perform its intended function.
Subject areas covered in this report include civil and structural and miscel-laneous issues. A report on the electrical and instrumentation and test pro-gram areas was published in January 1985. The technical concerns and allegations in the areas of mechanical and piping, coatings, and QA/QC, as well as the re-maining areas of outstanding regulatory actions, will be addressed in future supplements to the Comanche Peak Safety Evaluation Report (SER). The staff's findings for civil and structural and miscellaneous allegations or concerns are summarized in Section 3 of this Appendix. Attachment 1 to the Appendix is a listing of the technical concerns and allegations relating to civil and structural and miscellaneous issues. Details of the assessment and findings on individual concerns or allegations appear in Attachment 2 to this Appendix. Those aspects of the concerns or' allegations that pertain to wrong-doing (e.g., falsification of records) were forwarded to the NRC's Office of Investigations (01) for followup because they are outside the scope of the technical staff's review. A number of potential violations of NRC rules and regulations have been identi-fied during the course of the TRT investigation. These potential violations have not been addressed in this SSER, but will be further reviewed by the NRC Region IV staff, which will determine appropriate followup actions. Comanche Peak SSER 8 K-2
i l l i l i'
- 2. Comanche Peak Technical Concerns and Allegations Management Program
2.1 Background
i Shortly after the ED0's issuance of the March 12, 1984, directive, the staff found it necessary to (1) obtain current information relative to TUEC's manage-ment control of the construction, inspection, and test program and (2) obtain { necessary information to establish a management plan for resolution of all out-
- standing licensing actions. In order to achieve these goals in an expeditious '
i and objective manner, a Special Review Team (SRT) was formed to conduct an un-announced review of the Comanche Peak plant. The SRT consisted of eight j' reviewers and one team leader, all from NRC's Region II Office, and a team manager from NRC headquarters. The SRT spent over 800 manhours, from April 3 to 1
; April 13, 1984, performing this review. The SRT concluded that TUEC's programs were being sufficiently controlled to allow continued plant construction while 4
the NRC completed its review and inspection of the Comanche Peak facility. l The SRT review also provided a basis for the development of an NRC management i plan for the resolution of all outstanding licensing actions. This plan was
- approved on June 5,1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of
! NRC's Region IV Office. The purpose of the plan was to ensure the overall i
coordination and integration of the outstanding regulatory actions at Comanche
; Peak and their satisfactory resolution prior to a licensing decision by the NRC.
In accordance with the plan, a Technical-Review Team (TRT) was formed to evaluate and resolve technical issues and those allegations that had been identified. On July 9,1984, the TRT began its 10-week (five 2-week sessions) i onsite effort, including interviews of allegers and TUEC personnel, to determine the validity of the technical concerns and allegatiens, to evaluate their safety { ' significance, and to assess their generic implications. The TRT consisted of about 50 technical specialists from NRC headquarters, NRC Regional Offices, and NRC consultants, who were divided into groups according to technical discipline. Each group was also assigned a group leader. 2.2 Review Approach and Methodology 2.2.1 Concern and Allegation Tracking System. A tracking system was developed for identifying and listing each concern or allegation. These technical concerns and allegations were grouped according to
' their topical areas or disciplines, and were listed numerically within each group in the order that they were identified by the TRT. The tracking system ! included a description of the concern or allegation;-its status or the actions taken to resolve it; the nature of the sources of the concern or allegation (i.e., anonymous or confidential); a code for the individual who identified.the concern or allegation (instead of the individual's name); ,the date when the 4 concern or allegation was received by the TRT; the source document (e.g.,
4 Comanche Peak SSER 8 K-3
.____l_---_._._____ - _ _ _ _ _ _ _ . _ _ . . _ , _ - . _
letter, NRC inspection report, hearing transcript, etc.); cross reference; etc. At the end of each 2-week session, the concern / allegation tracking system was updated, as needed, to reflect the status of each concern or allegation, as well as any new ones that had been added. 2.2.2 Review Methodology The technical concerns or allegations similar in subject were combined and evaluated as one category. For each concern / allegation or concern / allegation category, an approach to resolution was developed by the cognizant reviewer (s). Each approach to resolution was reviewed and approved by the responsible group leader. The group leaders and reviewers were instructed to: develop and maintain a work package for each issue or category of issues that contained or referenced pertinent documentation associated with the issue (s) and the ultimate resolution, including records of interviews and inspections for supporting the final NRC staff decisions regarding the issue (s); and to protect the identity of the allegers, as a matter of NRC practice. Such efforts included limited and controlled distribution of allegation-related documentation and correspondence; minimal use of names, identifying titles, or position descriptions in written material; enlarged sampling of activ-ities to prevent direct links by non-NRC personnel between the activity under investigation and the alleger; and other indirect approaches toward investigating the allegations. During TRT onsite sessions, daily meetings were held at the review group level to assess progress, to adjust the inspection and evaluation approach as needed, and to provide a forum for the reviewers to interact with one another or to discuss problems and to arrive jointly at resolutions. Similar daily meetings were also held at the management level where the group leaders interacted with one another and with the Project Director, his assistant and staff. In evaluating the technical concerns and allegations, the TRT reviewers examined , areas in the plant where direct observation could provide information needed for - evaluating an allegation or concern. During its onsite sessions, the TRT inter-viewed the allegers as needed to clarify their concerns or allegations. To the extent possible, the TRT contacted allegers after its onsite review to discuss preliminary TRT findings and to obtain any additional comments from them. (See Section 2.2.3 below.) The TRT also interviewed TUEC and TUEC contractor per-sonnel as was warranted by the evaluation. In addition to these contacts, the TRT reviewed various project documents, including specifications, engineering drawings and analyses, procedures, instructions, NRC Region IV inspection 1 reports, and applicable sections of the Final Safety Analysis Report (FSAR) and NRC regulations pertinent to the allegation or sample selected by the TRT for i inspection. The TRT also examined construction records, such as design change authorizations, construction work packages, QC inspection reports, nonconform-ance reports, deficiency logs, lists and reports, and QC inspector training and certification records. In addition, the TRT reviewed pertinent transcripts from recent ASLB hearings and depositions of TUEC personnel and former employees. Comanche Peak SSER 8 K-4 t
,, . - - . ~ , .-- ..- _.
_m Based on these reviews and interviews, the TRT determined the validity of each technical concern or allegation and assessed its safety significance, its potential generic implications, and any indications of potential management breakdown. Detailed documentation of the TRT assessment and final determina-tions of each technical concern or allegation appear in Attachment 2 to this Appendix. 2.2.3 Interviews with Allegers Approximately 900 technical concerns and allegations regarding the construction of the Comanche Peak facility have been raised by approximately 70 allegers through various mechanisms. During its onsite work, the TRT interviewed 18 individuals in person, some of whom received followup interviews by telephone. For ten allegers, the TRT reviewers were able to obtain the needed information by telephone and determined that personal interviews would not be necessary. Three allegers contacted by the TRT declined being interviewed. Five allegers could not be located during the TRT's onsite sessions because their current addresses and telephone numbers were not available. They have not responded to correspondence from the TRT cent to their last known addresses expressing the TRT's intention to discuss their concerns with them. Efforts to locate these individuals included inquiries through the NRC's Office of Investigations, NRC's Region IV staff, the telephone company and U.S. Postal Service, selected inquiries of their relatives and former co-workers, confidential examination of the personnel files of TUEC and its contractors, and in some cases, inquiries to the intervenor group, the Citizens Association for Sound Energy (CASE), and the Government Acountability Project (GAP). To the extent possible, the TRT kept a transcript for each personal interview conducted during its onsite sessions. The names and identities of the allegers had been deleted from the transcripts, as well as from other pertinent reference or source documents, before TRT reviewers were given any portions of these documents for review and follow-up. During the TRT's onsite work, the original transcripts were kept in a locked file in the TRT Project Director's office. The distribution of these transcripts within the NRC, and even within the TRT, was limited and controlled. Subsequent to its onsite work, and at the completion of its evaluation, the TRT attempted to contact each alleger to discuss the TRT's findings regarding their original concerns, and to obtain additional comments from them, if any. Thirty allegers have received such followup interviews. A total of 19 allegers could not be located. Some of these individuals had received initial TRT interviews but had since left the area. Three allegers declined to have further contacts with the TRT. The TRT is in the process of contacting the remaining allegers for followup interviews. The outcome of followup interviews conducted through December 1984, is briefly discussed in the individual SSER sections in Attach-ment 2. Transcripts were kept for all followup interviews conducted either by telephone or in person. 2.3 Communications with TUEC Whenever the TRT reviewers encountered problems during their evaluations, the TRT Project Director and/or his designee resolved them through discussions with TUEC management onsite. There were also frequent staff-level contacts between TRT members and TUEC personnel during the TRT's onsite activities. In keeping Comanche Peak SSER 8 K-5
with the NRC practice of promptly notifying applicants of outstanding information/ evaluation needs that could potentially affect plant safety, the staff held several meetings with TUEC representatives at NRC headquarters toward the end of the TRT's review. These meetings were held to discuss potential safety concerns and to request additional information needed by the TRT to com-plete its review. The NRC staff met with TUEC representatives for the first of these meetings on September 18, 1984, to discuss TRT findings for electrical and instrumentation, civil and structural, and test program allegations and concerns. A letter docu-menting these findings and a request for additional information was issued to TUEC on the day of the meeting (Attachments 3 and 4). TUEC later submitted the requested information in the form of a proposed program plan, delineating planned actions to address the deficiencies identified by the TRT. The TRT met with TUEC representatives to discuss this proposed program plan on October 19 and 23, 1984. TUEC submitted a partially revised program plan to NRC on November 21, 1984. By letter dated January 24, 1985, the TRT provided TUEC with detailed comments on the program plan and issue specific action plans. On November 29, 1984, NRC sent a letter to TUEC containing potential open issues and requesting additional information and proposed program plans for mechanical and piping and miscellaneous allegations and concerns (Attachment 5). The letter also provided TUEC with the status of NRC's evaluation of coatings alle-gations. Informal telephone discussions between TRT group leaders and their TUEC counterparts regarding these letters have been ongoing. (Reports document-ing these discussions have been made available to CASE and are available for inspection at the NRC Public Document Room, 1717 H St., N.W., Washington, D.C. 20555, and at the Comanche Peak Local Public Document Room, Somervell County Public Library On The Square, P.O. Box 1417, Glen Rose, Texas 76043.) On January 8, 1985, the NRC issued a letter to TUEC informing them of the TRT's preliminary findings in the construction QA/QC area and requesting a program and schedule for completing a detailed and thorough assessment of the QA issues presented in the letter. A meeting between TUEC and the TRT was held on January 17, 1985, to discuss potential open issues in the QA/QC area. TUEC's proposed program plan for each of the subject areas and its implementation of the plan will be evaluated by the NRC staff prior to the NRC licensing decision on Comanche Peak. Comanche Peak SSER 8 K-6
- 3. Summary of Evaluations 3.1 Civil and Structural (C&S) Group Summary 3.1.1 Scope of Concerns and Allegations The concerns and allegations in the C&S discipline involved most aspects of reinforced concrete construction and testing. These allegations and concerns relate to (1) design deficiencies, (2) testing or inspection irregularities, (3) incorrect construction practices, (4) inadequate repairs, (5) uncorrected, unsafe conditions in the completed structures, and (6) premature structural loading. The total of 57 concerns and allegations were grouped by subject into the following 17 categories:
Category No. Subject Characterization of Concerns and Allegations 1 Inadequate Materials Used in Rejected aggregate was incor-Concrete porated in the basemat of Unit 1 reactor; unauthorized quanti-ties of water added to concrete used in basemat; rejected con-crete placed in turbine genera-tor building; concrete with excessive slump placed in con-tainment walls; concrete rejected for being over speci-fication limit on time to dis-charge was placed in the Circu-lating Water Intake Structure. 2 " Bad Concrete Work" and
" Sloppy" Placement of " Bad concrete work" and " slop-Concrete py" placement of concrete; place-ment of " soupy" concrete in a slab in the Auxiliary Building in the summer of 1976.
3 Placement of Concrete During Poor Weather Conditions Placement of concrete during rainstorm and without approval by QC personnel and during or immediately before freezing weather; some field-cured cylinders and standard-cured cylinders failed specification requirements for concrete strength, and the Schmidt rebound hammer test was misapplied. Comanche Peak SSER 8 K-7 l I
Category Characterization of Subject Concerns and Allegations No. 4 Concrete Void, Cracks, and Voids in concrete behind stain-Crumbling less steel liner of Unit 1 reac-tor cavity and in building walls; cracks in concrete basemat of Unit 1 and in floor slabs in the plant building; foreign material embedded in concrete; fresh concrete placed on top of crumbling concrete. 5 Miscellaneous Concrete Equipment was set on grout Construction Irregularities before the grout properly gained strength through aging; hanger inserts installed at improper angles; trash in bottom of a form was covered with concrete. 6 Rebar Improperly Installed Rebar was installed that was not or Omitted properly inspected upon receipt at the site; rebar omitted at various specified locations. 7 Uncontrolled Repair of Concrete A hole in a concrete slab result-ing from removal of a Hilti bolt in the floor of the Safeguards Building was repaired in an
" uncontrolled manner."
8 Falsification of Records Various specified records con-cerning concrete tests were falsified. 9 Improperly Conducted Inspector Inspector recertification tests Recertification Tests were done "open book" after March of 1977 and examinations were given with answers provided. 10 Violations of Testing Procedures Equipment required for aggre-gate testing unused; short cuts taken in aggregate testing; concrete placed without re-quired testing; concrete cylinder compression tests run at faster loading rate than permitted by NRC regulations; concrete test cylinders in the laboratory moist room allowed to dry. Comanche Peak SSER 8 K-8
Category Characterization of No. Subject Concerns and Allegations 11 Poor Workmanship in Use of Poor workmanship in use of Rotofoam rotofoam as a temporary spacer during construction to maintain required seismic gap between Category I concrete structures. 12 Concrete Construction A spillway pillar, span, or Deficiencies column was erected 75 to 80 degrees offset. 13 Concrete Cracks At Bottom Detrimental cracks in concrete of Reactor Vessel pad at bottom of reactor vessel. 14 Control Room Area Deficiencies The field run conduit, drywall, and lighting fixtures installed above ceiling panels in the control room are classified as nonseismic and are supported only by wires, and may fall as a result of a seismic event. 15 Unauthorized Cutting of Rebar Undocumented and unauthorized holes were drilled through rebar. 16 Excavation Overbreak/ Overexcavation and improper Seismic Response fill under Unit 1 Containment Building could invalidate expected seismic response of the foundation due to change in properties resulting from removal of in-situ material. 17 Improper Concrete Sampling Personnel produced incorrect readings on concrete batch plant scales by leaning on wires connecting the weighing hoppers to the scales. 3.1.2 Civil and Structural (C&S) Group The Civil and Structural Group consisted of three NRC employees and four consul-tants, all of whom are civil and structural engineers, with a' combined total of 137 years of experience in general design and in nuclear and non nuclear heavy construction work. These reviewers were selected for their technical expertise and experience in design, construction, quality assurance, and ability to detect discrepancies in construction records. Comanche Peak SSER 8 K-9
3.1.3 Findings for Civil and Structural Issues Fourteen of the 57 concerns and allegations reviewed by the TRT in the Civil and Structural area were not substantiated. Of the 20 that were substantiated, 3 were found to have potential safety significance. In addition, there are 2 allegations, although not substantiated, whose safety significance can not be determined at this point. TUEC has been requested to provide more information to the NRC staff before these issues can be resolved. The TRT could not determine the validity of 21 allegations. However, a conservative approach was taken to disposition the allegation, i.e., the TRT assessed the potential structural significance of the allegation assuming that it was true. Two allegations simply reiterated allegations already made. The first issue that has been substantiated and was determined to be of poten-tial safety significance involved reinforcing steel (rebar) omitted from a con-crete placement in the reactor cavity wall of Unit 1. The C&S Group requested documentation indicating that an analysis was performed supporting the omission of this rebar. The C&S group was subsequently informed that an analysis had not been performed. Therefore, the safety significance of this issue cannot be determined until an analysis is performed verifying the adequacy of the reinforcing steel as installed. (See Attachment 2, C&S Category 6.) Another issue that has been substantiated and was determined to be of potential safety significance concerned the maintenance of an air gap between concrete structures. Based on a review of available inspection reports and related documents, on field observations, and on discussions with TUEC' engineers, the C&S Group could not determine if an adequate air gap had been provided between concrete structures. In addition, it is not apparent that the permanent installation of elastic joint filler material (rotofoam) between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the assumptions and dynamic models used to analyze the buildings. (See Attachment 2, C&S Category 11.) The third issue that has been substantiated and was determined to be of poten-tial safety significance concerned the seismic design of the control room ceiling elements. This issue was jointly reviewed by_the C&S Group and the Electrical and Instrumentation Group. For the nonseismic items (other than the sloping suspended drywall ceiling), and for nonsafety-related conduits whose diameter is 2 inches or less, the C&S Group could find no evidence that the possible effects of a failure of these items had been considered. In addition, the C&S Group determined that calculations for seismic Category II components (e.g., lighting fixtures) and the calculations for the sloping suspended drywall ceiling did not adequately reflect the rotational interaction with the nonseismic items. The fundamental frequencies of the supported masses had not been determined to assess the influence of the seismic response spectrum at the control room ceiling elevation on the seismic response of the ceiling elements. (See Attachment 2, C&S Category 14.) The C&S group investigated the technical implications concerning falsification i of concrete strength tests. The preponderance of evidence suggests that falsi-fication did not occur. However, since a number of other allegations were resolved on the basis of concrete strength results, the C&S Group believes that action is required on the part of TUEC to provide confirmatory evidence that the reported concrete strength test results are indeed representative of the 1 Comanche Peak SSER 8 K-10
- - ~ _ . . - - - ~_..-- -. .- - _- . -- .. . . . - _-
i i strength of the_ concrete placed in Category I concrete structures. (See
- Attachment 2, C&S Category 8.)
4
- Another issue that was not substantiated and whose safety significance could not yet be determined concerned the unauthorized cutting of rebar in the Fuel
- Handling Building. The C&S Group found that if certain holes were drilled to i the depth alleged, rebar would have been cut without authorization. (See Attachment 2, C&S Category 15.)
; The C&S Group found that the allegation concerning hollow places in concrete-behind the stainless steel liner of the Unit 2 Reactor Cavity is true; the hollow places are currently undergoing repairs. The repairs and the repair documentation must be inspected, reviewed, and approved by the NRC before the j TRT can determine that this issue has been adequately resolved. (See Attach- ; ment 2, C&S Category 4.)
i The C&S Group could not substantiate the concerns raised by the remaining alle-
! gations and concluded that these concerns have no structural safety significance.
- However, the results of the evaluations for Categories 1,3,4,5,6,7,8,9, i 10, 15, and 17 are being further assessed by the QA/QC Group as part of its
- overall programmatic review. (See Attachment 2, C&S Categories 1, 3, 4, 5, 6,
] 7, 8, 9, 10, 15 and 17.) 3.1.4 Overall Assessment and Conclusions
- During its evaluations, the TRT reviewed pertinent construction records (e.g. ,
t concrete placement packages, NCRs, concrete test results), structural design drawings and calculations, specifications (e.g. , for concrete reinforcing steel), interviewed craft and TUEC personnel and conducted plant inspections. This documentation, to the extent reviewed by the TRT, was judged to be ade-j quate and consistent with applicable FSAR' commitments, except for the
! deficiencies identified in the SSER sections in Attachment 2. Therefore, the i
TRT concludes that the civil and structural construction within the scope of j the TRT C&S group review effort was adequate and was, for the most part, well documented. i Five issues in the civil and structural area still require further action. One case involving reinforcing steel omitted from the-reactor cavity wall, and t another case of alleged unauthorized drilling of reinforcing steel, require further documentation. TUEC must also test concrete in place to evaluate an i allegation concerning falsified concrete strength tests. In addition, TUEC i must conduct analyses and inspections to determine whether the separation ! between buildings is adequate to provide acceptable performance in an earthquake. j Finally, there must be a seismic analysis of the suspended ceiling, lighting , j fixture and nonsafety-related conduit in the control room to demonstrate design i adequacy of the ceiling elements. The potential safety-implications of this i'
- issue for nonseismic structures, systems, and components in_other parts of the
! plant must also be evaluated. { 3.2 Miscellaneous Group Summary i 3.2.1 Scope of Concerns and Allegations , The allegations with a Miscellaneous designation covered a wide. variety of t topics and involved both administrative and construction activities, some Comanche Peak SSER 8 K-11
q 1 safety related and some nonsafety related. In total, 29 allegations were designated as Miscellaneous; i.e., their subject matter did not fall within the scope of responsibility of one of the other Technical Review Team's technical disciplines. Twenty-five allegations were subsequently consolidated into 20 categories, each of which dealt with a general topic; three were transferred to the TRT Mechanical and Piping group for review and followup and one to the Office of Investigations. The following is a listing and description of each of the 20 Miscellaneous categories. Category Characterization of Concerns and Allegations No. Subject Nuclear Fuel Nuclear fuel was received prior to issuance of 1 special nuclear material license. 2 Reactor Pressure Expansion of the RPV during hot functional test-Vessel (RPV) ing caused the vessel reflective insulation to come in contact with the concrete biological shield wall; the Unit 1 RPV is located 3/16 inch off center. 3 Comanche Peak The Comanche Peak PSAR contains errors. PSAR 4 Radioactive Radioactive material was thrown into the lake. Material thrown into Comanche Peak Reservoir 5 High Pressure Cracks were observed in lower casing of the high Turbine pressure turbine. 6 Pressurizer Area A section was cut from a prefabricated pipe Piping "in the pressurizer area." 7 Unit i Main Design and fabrication problems were associated Condenser with the main condenser. Component Cooling Anchor bolts were damaged during installation. 8 Water Surge Tank 9 Hayward Tyler Hayward Tyler pumps in safety systems may have Pump Deficiencies unidentified deficiencies because of a poor quality assurance program at Hayward Tyler. Unit 1 Diesel Two Unit 1 diesel generators we e damaged. 10 Generators 11 Polar Crane Shimming and installation of the polar crane were improper. A deficient weld on a door was accepted. 12 Missile Barrier Door K-12 Comanche Peak SSER 8
Category No. Subject Characterization of Concerns and Allegations 13 Tube to Base Tube steel was cut at the wrong angle and welded Plate Weldments to a baseplate, leaving a large gap between the tube and baseplate. 14 NRC Form-3 NRC Form-3 was posted at an insufficient number Posting of site locations. 15 Orug Abuse Drug use and abuse was widespread and management did not give proper attention to the alleged problem. 16 HVAC Heating, ventilating, and air conditioning system (HVAC) supports for seismic loads were not ana-lyzed; HVAC components and supports inside con-tainment were not properly considered as missiles; HVAC failure during a postulated accident would allow temperatures to rise to an unacceptable level inside containment. 17 Reactor Vessel Damage occurred to upper 1cternals of the reactor Internals vessel. 18 Polar Crane Internal wires were broken in the polar crane Cables festooned cables. 19 Radwaste System Workers habitually urinated on stainless steel Contamination pipe. 20 Instructions to Inadequate rigging and handling instructions were Craft Personnel provided to craft personnel. 3.2.2 Miscellaneous Group The members of the Miscellaneous Group were assembled based on their technical expertise, capabilities, and experience in engineering design, quality assur-ance and document control, inspection, construction, and regulatory activities. The group included five members from NRC's Region IV office, with expertise in various technical disciplines, and three consultants. Collectively, the group possessed experience in excess of 50 years in the nuclear power industry and its regulation. 3.2.3 Findings for Miscellaneous Issues Fourteen of the 24 allegations (grouped into 20 categories) pertained to systems and components classified as nonnuclear safety.(NNS) in the Comanche Peak Steam Electric Station (CPSES) Final Safety Analysis Report (FSAR). These , allegations (AM-3, 4, 5, 6, 9, 15, 16, 17, 22, 24, 25, and 30) were also not l listed as quality assurance (i.e., safety related) items in Table 17A of l Volume XIV of the FSAR. Accordingly, 10 CFR Part 50 Appendix B quality assurance requirements would not apply _to these systems and components except I l Comanche Peak SSER 8 K-13 l L
f 4-i for seismic considerations. Seven of the 14 al % ations have seismic classifi-cations. Ordinarily the NRC does not inspect items that are classified NNS or ; nonsafety related but would observe and bring deficient non-Q items to the , attention of Texas Utilities Electric Company (TUEC) for resolution. However, i these items were inspected by the Technical Review Team (TRT) because the TRT was responsible for resolving all allegations and assuring that nonsafety issues did not have safety implications. Four of the 14 allegations were ; potentially safety significant and potentially had generic implications; however, TUEC had identified and corrected the problems concerning part of . AM-25 (crane movement) and AM-30. Only AM-15 and 16 (Polar Crane Shimming) [' and AM-3 (Reactor Pressure Vessel Reflective Insulation) remain unresolved at this time and are identified as the first and second issues in the following l paragraph. i Ten of the 24 allegations (AM-2, 7, 12, 13, 14, 18, 19, 21, 23(a) and 23(b))
~
pertained to matters or systems which are classified as safety related. Four of these 10 allegations (AM-13, 14, 21 and 23(a)) were potentially safety i significant and had generic implications. However, TUEC had identified and l corrected (or was in the process of correcting) problems described in ! allegations. l t The first issue having potential safety significance (AM-3) involved the gap l between the reactor pressure vessel insulation and the biological shield wall. i Investigation of the allegation that the Unit I reactor pressure ~ vessel outer j wall was touching the concrete biological shield wall indicated that this j allegation was not factual. However, a significant construction deficiency t report documented that unacceptable cooling occurred in the annulus between the ' reactor pressure vessel reflective insulation (RPVRI) and the shield wall ' during hot functional testing, apparently because of the existence of an i inadequately sized annulus gap and possibly because of the presence of con- t struction debris in the annulus. TUEC corrected the situation by modifications ; to allow increased air flow for proper heat dissipation and by removal of the r construction debris. TUEC representatives indicated that testing to verify the , adequacy of the cooling flow will take place when additional hot functional i testing is conducted. Information gathered during the inve'stigation indicated ! that a design change in the RPVRI support ring (i.e., locating the ring outside l l rather than inside the insulation) resulted in a limited clearance between the . RPVRI and the shield wall. However, TUEC failed to: (1) address the funda-mental issue of the design change impact on annulus cooling flow, and (2) deter- ; mine whether Unit 2 was similarly affected. Consequently, further action is i required. (See Attachment 2, Miscellaneous Category 2.) The second issue-having potential. safety significance (AM-15 and 16) involves the polar crane rail support system. The installation of the polar crane rail ; support system was investigated by visual inspection, review of associated ! documentation, and discussions with TUEC representatives and their contractors. i Region IV documented that gaps on the Unit 1 polar crane bracket and seismic l connections exceeded design requirements. In TUEC responses, the gaps were t attributed to crane and bolting self-adjustment resulting from crane operation. A site design change was issued to document the acceptability of the gaps in excess of 1/16 inch which were identified in the NRC inspection report. ; 1 ! ! t i _ Comanche Peak SSER 8 K-14
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During further investigation of the allegation that shims for the rail support system of the polar crane had been altered during installation, gaps, which may have been excessive, were observed between the crane girder and the girder sup-port bracket. Detailed specifications addressing the gap tolerance in the girder seat connections did not exist; however, Gibbs & Hill indicated in a November 28, 1977 letter (GHF-2207) that seated connections do not require shimming, since the area in bearing is at least the width of the bottom flange of the crane girder. Contrary to this assumption, nine girders were observed to have gaps which extended under the bottom flange that reduced the bearing surface to less than the 20-inch flange width stated in the letter. The TRT also observed conditions which indicated that the crane rail may still be moving in a circumferential direction, that three rail-to-rail ground wires were broken, that two shims have partially worked out from under the rail, and that two Cad-welds were broken. (See Attachment 2, Miscellaneous Category 11.) The TRT found that 21 of 24 allegations (that is, all except AM-3, 15 and 16) were either unfounded or involved nonsafety-related issues, or the deficiency was identified by TUEC's quality assurance / quality control program and correc-tive actions had been completed that were acceptable to the TRT. 3.2.4 Overall Assessment and Conclusions The TRT found that 9 of the 24 allegations were substantiated, were potentially safety significant, and had generic implications. However, actions taken because of NRC Bulletins, inspections, and TUEC audits / evaluations corrected all but two problems. Therefore, the TRT concludes that 21 of 24 allegations had neither safety significance nor generic implications. The two problems for which TUEC will have to complete actions and address issues are Miscellaneous Category 2, the gap between the reactor pressure vessel reflective insulation and the biological shield wall, and Miscellaneous Category 11, improper shimming and installation of the polar crane rail support system. (See Section 4.2.) Once these actions are satisfactorily completed by TUEC and are reviewed and accepted by the NRC, a finding can then be made that no outstanding issues raised by the miscellaneous concerns and allegations remain that would preclude licensing of CPSES Unit 1. Comanche Peak SSER 8 K-15
- 4. Actions Required of TUEC TUEC shall submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified in the following sections. This program plan and its imple-mentation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic implications on safety-related systems, programs, or areas. The collective significance of these deficiencies should also be addressed. The program plan should also include'the proposed TUEC action to ensure that such problems will be precluded from occurring in the future. The specific actions required of TUEC are described in the following sections.
4.1 Civil and Structural (C&S) Area 4.1.1 Rebar Improperly Installed or Omitted (See Attachment 2, C&S Category 6)
- Provide an analysis of the as-built condition of the Unit 1 reactor cavity that verifies the adequacy of the reinforcing steel between the 812-foot and 819-foot, -inch elevations. The analysis shall consider all required load combinations.
4.1.2 Falsification of Concrete Compression Strength Test Results (See Attach-ment 2, C&S Category 8)
- Determine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a program to assure acceptable con-crete strength. The program shall include tests such as the use of random Schmidt hammer tests on the concrete in areas where^ safety is critical.
The program shall include a comparison of the results with the resuits of tests performed on concrete of the same design strength in areas where the strength of the concrete is not questioned to determine if any signifi-cant variance in strength occurs. TUEC shall submit the program for these tests to the NRC for review and approval prior to performing the tests. 4.1.3 Maintenance of Air Gap Between Concrete Structures (See Attachment 2, C&S Category 11)
- Perform an inspection of the as-built condition to confirm that adequate separation for all seismic Category I structures has been provided.
- Provide the results of analyses which demonstrate that the presence of rotofoam and other debris between all concrete structures (as determined by inspections of the as-built conditions) does not result in any signi-ficant increase in seismic response or alter the dynamic response charac-teristics of the Category I structures, components, and piping when com- ,
l pared with the results of the original analyses. Comanche Peak SSER 8 K-16
- -- - =- - . _ . 4.1.4 Seismic Design of Control Room Ceiling Elements (See Attachment 2, C&S Category 14) Provide the results of seismic analysis which demonstrate that the non-seismic items in the control room (other than the sloping suspended dry-wall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8. Provide an evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceil-ing (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems. Verify that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy appli-cable design requirements. Provide the results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit 2 inches or less in diameter. Provide the results of an analysis which demonstrate that the foregoing problems are not applicable to other category II and nonseismic structures, systems, and components elsewhere in the plant. 4.1.5 Unauthorized Cutting of Rebar in the Fuel Handling Building (See Attach-ment 2, C&S Category 15) Provide information to demonstrate that only the No. 18 reinforcing steel in the first layer of the floor slab at the 810-ft, 6-inch elevation of the Fuel Handling Building was cut during installation of the trolley process aisle rails, or Provide design calculations to demonstrate that structural integrity is maintained if the No. 18 reinforcing steel on both the first and third layers of the floor slab was cut. 4.1.6 Hollow Places in Concrete Behind Unit 2 Reactor Cavity Liner (See Attachment 2, C&S Category 4) Provide details of the successful completion of the repairs to the hollow places in concrete behind the Unit 2 reactor cavity liner. 4.2 Miscellaneous Area 4.2.1 Gap Between Reactor Pressure Vessel Reflective Insulation (RPVRI) and the Biological Shield Wall (See Attachment 2, Miscellaneous Category 2)
- Review the procedures for approval of design changes to non-nuclear safety-related equipment, such as the RPVRI, and make revisions as necessary to assure that such design changes do not adversely affect safety-related systems.
Comanche Peak SSER 8 K-17
- Review procedures for reporting significant design and construction defi-ciencies, pursuant to 10 CFR Part 50.55(e), and make changes as necessary to assure that complete evaluations are conducted. - Provide an analysis which verifies that the cooling flow in the annulus between the RPVRI and the shield wall of Unit 2 is adequate for the as-built condition. - Verify during Unit 1 hot functional testing that completed modifications to the RPVRI support ring now allow adequate cooling air flow.*
4.2.2 Control of Debris in Critical Spaces Between Components and/or Structures (See Attachment 2, Miscellaneous Category 2; also see Attachment 2, C&S Category 11)
- Identify areas in the plant having critical spacing between components and/or structures that are necessary for proper functioning of safety-related components, systems, or structures in which unwanted debris may collect and be undetected or be difficult to remove; - Prior to fuel load, inspect the areas and spaces identified and remove debris; and - Subsequent to fuel load, institute a program to minimize the collection of debris in critical spaces and periodically inspect these spaces and remove any debris which may be present'.
4.2.3 Polar Crane Shimming (See Attachment 2, Miscellaneous Category 11)
- Inspect the polar crane rail girder seat connections for the presence of gaps which reduce the bearing surface to less than the width of the bottom flange, and perform an analysis which will determine whether existing gaps are acceptable or require corrective action.
- Determine if additional rail movement is occurring and, if so, provide an evaluation of safety significance and the need for corrective action.
- Perform a general inspection of the polar crane rail and rail support system, correct identified deficiencies of safety significance, and pro-vide an assessment of the adequacy of existing maintenance and surveill-ance programs.
*This testing has been completed. However, TUEC's analysis of the test results is still underway.
Comanche Peak SSER 8 K-18
ATTACHMENT 1 LISTING OF TECHNICAL CONCERNS AND ALLEGATIONS I. Civil and Structural Allegation Number Characterization Category Page Number AQC-1 Concrete air entrainment test records 8 K-59 were falsified. AQC-2 Concrete laboratory test records were 8 K-59 falsified. AQC-3 Concrete aggregate tests were falsified. 8 K-59 AQC-4 Equipment required for aggregate testing 10 K-71 had not been used. AQC-5 Improper methods were used to dry coarse 10 K-71 aggregate for sieve analysis. AQC-6 Some of the Unit 1 Containment Building 10 K-71 basemat concrete was placed without i required testing. AQC-7 Concrete compressive strength test 8 K-59 results were falsified. AQC-8 Concrete compressive strength test speci- 10 K-71 mens were loaded at an excessive rate. AQC-9 Recertification examinations for 9 K-67 R. W. Hunt inspectors were given open book and examinations were given with answers supplied. AC-10 Concrete repair following removal of a 7 K-57 Hilti bolt was improper. AQC-11 Acceptable concrete test cylinders were 10 K-71 used to represent concrete placements other than those for which the samples were made. AQC-12 Reinforcing steel (rebar) was installed in 6 K-49 the Containment Building without quality control (QC) inspection. K-19 L _- . - . - .. . - - . .- .
l l l l I. Ciyil avid Structurel (Continued) Allegation . Page Number Characterization Category Number AC-13 Diamond core drill bits were loaned for 15 K-87 the unauthorized and undocumented cutting of rebar. AC-14 There was unauthorized cutting of rebar 15 K-87 in nonspecific locations. AC-15 There was unauthorized cutting of rebar 15 K-87 during installation of trolley process aisle rails in the Fuel Handling Building. Rejected aggregate was used in the Unit 1 1 K-27 AQC-16 Reactor Building basemat. AC-17 (This allegation was reassigned to the Mechanical and Piping Group and is assessed in Mechanical and Piping Category 13,
" Metal Shavings from Drilling Fuel Pool Underwater Lamps," under Allegation AP-7.)
AE-17 Field run conduit, drywall, and light- 14 K-83 ing installed above the control room were classified nonseismic and were thus inadequately supported. AC-18 There was unauthorized cutting of rebar 15 K-87 in nonspecific locations. AC-19 Truck drivers added unauthorized quanti- 1 K-27 ties of water to concrete used in the basemat. AC-20 Rejected concrete was placed in the Tur- 1 K-27 bine Building. AC-21 A batch of concrete with excessive slump 1 K-27 was placed in the containment structure. AC-22 " Bad concrete work" and " sloppy" placement 2 K-31 of concrete occurred in unspecified locations. AC-23 See AC-22. - 2 K-31 AC-24 A batch of concrete was placed in the 3 K-33 Containment Building dome during a rain-storm without the presence of a QC inspector. K-20
I. Civil and Structural (Continued) Allegation Number Characterization Category Page Number AC-25 Voids existed in the concrete wall behind 4 K-39 the Unit 1 reactor cavity stainless steel liner. AC-26 Equipment was set on grout before the 5 K-45 grout properly gained its required strength through curing. AC-27 Rejected and improper material was used 1 K-27 in concrete batches. (See AQC-16, AC-19, AC-20, AC-21, AC-47.) AC-28 Fresh concrete was placed on top of 4 K-39 crumbling concrete during construction of a spillway.
; AC-29 A pillar, span, or column associated with 12 K-79 1
a spillway was erected 75 to 80 degrees offset. AC-30 Rebar was omitted from a portion of the 6 K-49 Safeguards Building 1 AC-31 Richmond Insert anchor bolt inserts were 5 K-45 installed in Unit 1 at angles not per-pendicular to the concrete surface. f AC-32 A 20-ft by 20-ft area of honeycombed con- 4 K-39 1 crete in the Unit 1 Auxiliary Building was inadequately repaired. AC-33 Cracks exist in the Unit 1 concrete 4 K-39 i basemat and in the floors of other i plant buildings. l AC-34 Concrete voids could be detected in 4 K-39 building walls by tapping with a hammer and-listening for a hollow sound. AC-35 Concrete was placed in the Safeguards 3 K-33 i Building basemat and.the lowest level floor of the Unit 1 Containment Building during or just before freezing weather. AC-36 Trash was placed in a form and then 5 K-45 covered with concrete. i K-21 1 _.- - . , - - , , . . , ,, -._..,,e - . , _ -. - - , --, r, , ,_,, ,. , . -,, - . - . -.,-._.m_ , - , ,..-,._-,# . . , _ _ , , .
l l i l l r ( I. Civil and Structural (Continued) , Allegation Number Characterization Category Page Number AC-37 Rebar used in the containment structure 6 K-49 was not properly inspected upon its receipt at the site. (See AQC-12.) AC-38 Horizontal tie rebar was missing from the 6 K-49 ; i Unit 1 Containment Building wall. AC-39 Rebar was missing from four column faces 6 K-49 l along column line EA of the Auxiliary ! Building at-the 807-ft elevation. ! AC-40 There was unauthorized cutting of rebar 15 K-87 ! in nonspecific locations. 4 AC-41 There was poor workmanship in the use 11 K-75
- of elastic joint filler material, 1 "rotofoam," as a temporary spacer in l order to achieve the required airspace between seismic Category I structures.
j AC-42 (This allegation is a duplication of i allegation AQ-10, " Falsification of Civil- ! QC Records," and has been forwarded to the NRC Office of Investigations (OI) for
; followup.) . AC-43 (This allegation reiterated the concerns 5 K-45 ! in AC-26, AC-31, and AC-36.)
, AC-44 Cracks existed in the concrete' pad beneath 13 K-81
-the reactor vessel.
AQC-45 Somebody produced incorrect scale readings 17- K-95 at the concrete baten plant by leaning on the wires connecting the weight hoppers to the scales. h I AQC-46 Midpour test records associated with the. 8 K-59 Unit 1 Containment Building basemat were l falsified. , AC-47 Concrete rejected for being over speci- 1 'K-27 fication on time to discharge was placed in the Circulating Water Intake Structure. j AQC-48 Concrete test cylinders in the R.' W. H'nt- u 10 K-71
! laboratory moist room were allowed to dry.
K-22 i _ . _ _ _ - . , . - ,. . . ~ . _ . _ _ - _ . . _ , . - . . . . _ . , _ _ . . _ _ . _ _ . _ . _ , _ . _ _ , - . _ , _ _
4
- I.
Civil and Structural (Continued) Allegation Number Characterization Category Page Number AC-49 Rebar was installed upside down in a build- 6 K-49
- ing near the Unit 2 containment structure.
AC-50 " Soupy" concrete was placed in a slab in 2 K-31 i the Auxiliary Building during the summer of 1976. AQC-51 Cadweld tensile test results were 8 K-59 recorded during the spring and summer of 1976 without the tests having been j performed.
- AC-52 Several examples of field-cured cylin-
+ 3 K-33 ers and standard-cured cylinders failed specification requirements. The Schmidt rebound hammer test was then misapplied ] to resolve these test failures. AQ-64 Overexcavation and improper fill under 16 K-93 the Unit I containment structure could
' invalidate the expected seismic response of the foundation due to changes in properties from the removal of in-situ material.
i j II. Miscellaneous
; AM-1 (Issues from this allegation were addressed by the Electrical Group
[AE-50 and AE-51]; the Mechanical and Piping Group [AP-24, AP-25, AP-26,AP-27,AP-28,AQW-69,AQW-71]; and the QA/QC Group [AQ-111].) i i AM-2 Nuclear fuel was received onsite 1 K-97 before the NRC issued a special. j materials license. ! AM-3 During hot functional testing, 2 K-99 expansion caused the reactor pressure vessel reflective insulation to touch the biological shield wall. AM-4 The Preliminary Safety Analysis Report, i 3 K-103 Sections 10.2-11 and 10.2-12, contained errors. K-23 1
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II. Miscellaneous (Continued) Allegation Characterization Category Page Number Number AM-5 There was a possibility that someone 4 K-105 threw radioactive material into the Comanche Peak reservoir. AM-6 There were cracks in the lower casing 5 K-107 of the high pressure turbine. AM-7 A section of prefabricated pipe was 6 K-109 cut "frcm the pressurizer area." AM-8 The Unit 1 main condenser tubes were 7 K-111 beaten with hammers, were split during belling and flaring, and were improperly rolled. AM-9 The condenser tube support sheets had 7 K-111 holes that were misaligned by 3/8 inch. AM-10 The turbine-to-condenser tubing was mis- 7 K-111 aligned and then jacked into alignment causing stress. AM-11 (This allegation was transferred to the Mechanical and Piping Group, Category 43.) AM-12 The anchor bolts were damaged during the 8 K-115 installation of the component cooling water surge tank. AM-13 Pumps manufacturered by the Hayward Tyler 9 K-117 Pump Company were installed in Comanche Peak safety systems. These pumps may have unidentified deficiencies because of the poor QA program at Hayward Tyler. One of the diesel generators was 10 K-119 AM-16 damaged in May 1982. Shims for the rail suppport system 11 K-121 AM-15 for the polar crane were altered during installation. The polar crane rail moves during 11 K-121 AM-16 crane operation such that large gaps develop. Deficient welds on a missile barrier 12 K-125 AM-17 door were accepted. K-24
l l l l II. Miscellaneous (Continued) ! Allegation 4 Number Characterization Category Page Number , i AM-18 The tube steel used to fabricate 13 K-127 l supports in the Unit 1 safeguards
! "796 yard tunnel" was cut at the 1 wrong angle, resulting in excessive 4 gaps for the weld joints between the tube steel and baseplates.
AM-19 The posting requirements for NRC 14 K-131
- Form 3 were not met from 1977-1982.
i AM-20 Material false statements were made by plant management to the Atomic Safety and Licensing Board. (This allegation was transferred to the NRC Office of Investigations for followup.) AM-21 There was widespread drug abuse 15 K-133 at Comanche Peak, and management did not give proper _ attention to this problem. AM-22 TUEC has not analyzed the heating, 16 K-137
! ventilating, and air conditioning system (HVAC) supports for seismic loads. HVAC components and sup-ports inside containment were not properly considered as missiles.
! HVAC failure during a postulated accident would allow temperatures to rise to an unacceptable level inside ! containment. AM-23(a) A craft person stated that he had 20 K-147 4 not received instructions about.how to rig and handle a large motor-operated valve. AM-23(b) The Unit 1 reactor pressure vessel 2 K-99 ) is located 3/16 inch west of the north-south centerline through the containment building. AM-24 15-foot by 2 -inch stainless steel 17 K-139 bars inside the Unit 1 reactor e vessel upper internals were damaged I and then repaired without proper
- documentation.
l ! -K-25 i
II. Miscellaneous (Continued) l Allegation Number Characterization Category Page Number AM-25 Internal wires were broken in the 18 K-143 polar crane festooned cables, and the polar crane hit some hangers while operating. AM-26 (This allegation was transferred to the NRC Office of Investigations (0I) for followup) AM-27 (This allegation was transferred to OI for followup) AM-28 (This allegation is the same as AM-29 below, and was transferred to the Mechanical and Piping Group, Category 39.) AM-29 (This allegation was transferred to the Mechanical and Piping Group, i Category 39.) AM-30 Workers habitually urinated on 19 K-145 stainless steel pipe located in the radwaste system. AM-31 (This allegation was tranferred to the Mechanical and Piping Group, Category 49.) K-26
ATTACHMENT 2 ASSESSMENT OF INDIVIDUAL TECHNICAL CONCERNS AND ALLEGATIONS IN CIVIL AND STRUCTURAL AND MISCELLANEOUS AREAS
- 1. Allegation Category: Civil and Structural 1, Inadequate Materials Used in Concrete
- 2. Allegation Number: AQC-16, AC-19, AC-20, AC-21, AC-27 and AC-47
- 3. Characterization: It is alleged that the following violations of specifi-cations occurred at various times:
- a. Rejected aggregate was incorporated in the basemat of the Unit I reactor (AQC-16).
- b. Truck drivers added unauthorized quantities of water to concrete used in the basemat (AC-19).
- c. Rejected concrete was placed in the turbine generator building (AC-20).
- d. Concrete with excessive slump was placed in containment walls (AC-21).
- e. Some concrete was placed in the Circulating Water Intake Structure after the concrete was rejected for being over specification limit on time to discharge (AC-47).
AC-27 contained no new allegations; it merely reiterated those already made. Allegations AC-19, AC-20, and AC-21 were investigated by Region IV and documented in inspection report 79-09, which was reviewed by the NRC Tech-nical Review Team (TRT) as a step in its own assessment of the allegations.
- 4. Assessment of Safety Significance: Allegations AC-19, AC-20, and AC-21 appeared in a newspaper article. The identify of the alleger of AC-19 was not disclosed and therefore could not be contacted. Allegations AQC-16 and AC-47 were judged as having sufficient clarity for technical resolu-tion without initial contact between the TRT and the allegers.
- a. The TRT cannot determine whether or not the allegation that " rejected" aggregate was used is valid (AQC-16). The only item in the record was Deficiency and Disposition Report C-446 (December 9, 1976), which stated that an untested pile of aggregate, rather than an unacceptable pile, was used. Therefore, the acceptability of the aggregate is unknown. The alleger also stated that the equipment operator scraped aggregate off the floor of the storage area and dumped it on the conveyer belt so that it bypassed testing. The consequences of this alleged action may be evaluated by the effect on the properties of fresh and hardened concrete. The purpose of controlling aggregate grading is to maintain concrete of uniform workability and strength.
A TRT examination of the concrete basemat placement record packages revealed that workability and strength were satisfactory throughout the placement. Less than 3 percent of the concrete was rejected for improper slump, and all concrete tested met the specifications for K-27
compressive strength. If any aggregate did not comply with specified grading, the deficiency did not materially affect the concrete prop-erties of the basemat. i
- b. Construction Procedure CCP 10, para. 4.10.5.6, required the signature
- of a representative from both the contractor and testing laboratory
' when water was added to the concrete after it left the plant (AC-19).
l The batch weights were such that it was possible to add some water to j the batch in the ready-mix truck without exceeding the maximum
- permitted water-cement ratio. The amount of the addition permitted i was printed on the batch ticket. However, before the addition was made, the written permission of the testing laboratory was required.
The TRT examined all 268 batch tickets and discovered in Concrete Placement Package 101-2781-001, 7-17-75, that 7 batches had water added. For these tickets, the only signature filled in belonged to the contractor representative; none of these tickets was signed by l the test laboratory representative. In each case, the volume of water added was within the range permitted. Although the contractor erred in not getting test laboratory approval, the additions should have had no adverse effect on the concrete. This error, however, indicated a breakdown in the quality control system. A TRT examina-tion of test results indicated that all were within specification guidelines: slump values ranged from 1 inch to 2-3/4 inches; air content was from 2.0 percent to 3.2 percent; and 28-day compressive strengths ranged from 5340 psi to 6671 psi. In addition, the TRT examined parts of the basemat which were still accessible. While only a small portion could be examined visually and this portion did not necessarily include any batches with added water, the portion examined was in excellent condition.
- c. Tte alleger did not indicate where in the turbine building the alleged infraction occurred (AC-20). The building contains over 700 l
small concrete placements, and all were available for examination. The TRT examined a random selection of 65 concrete placement packages and found no irregularities. However, the turbine generator building is a nonsafety-related structure. The Final Safety Analysis Report, Sec. 3.2, " Classification of Structures, Components, and Systems,"
- indicates that the turbine generator building is not a seismic Category I structure. Its structural failure would not affect safety i during a safe shutdown earthquake; therefore, the activity alleged to have occurred would not affect the safety of the plant.
l d. The alleger claimed that a batch with a slump of 4-1/4 inches was placed (AC-21). The slump requirement-in Gibbs & Hill (G&H) Specification 2323-SS-9, Revision 4,;Section 5.2, states: A tolerance of up to 1 inch above the indicated maximum shall be i allowed for individual batches provided the average of all batches tested or the most recent 10 batches tested, whichever , is fewer, does not exceed the maximum limit, i.e., 4 inches. Whenever the measured slump exceeds the indicated maximum by more than 1/4 inch, successive batches or truck loads as i deposited shall be measured until the slump is within the maximum limit. K-28 ,
. _ - . _ - , - - - - , - - - - - , - - - . - - . . - . , - c,--- . - . - , , - - - , _ . , . . - , . _ , - - -
W i 1 ! Thus, placement of a batch with a 4\-inch slump was permitted as long as.the average of all-batches or the nost recent 10 batches did not. exceed 4 inches and individual batches'did not exceed 5 inches. A single high slump batch, provided the slump does not exceed 5 inches, cannot constitute a violation of specifications.
- e. The TRT reviewed 51 (37%) concrete packages out of the 140 concrete packages for the Circulating Water Intake Structure which is a non-safety-related structure (FSAR Vol. IV, Sec. 3.2) (AC-47). Of the 51 reviewed, 13 batches of concret.e were rejected, 9 for test failure (air, slump, temperature), and 4 for being over the speci+ication limit on time to discharge. !t was noted on the batch ticket of each rejected batch of concrete where the concrete was dumped. None of the rejected batches was placed in the circulating water intake; they were placed in temrorary sl' abs which the contractor was placing at j
t the time. '
- 5. Conclusion and Staff Positions: The allegations were found to have no structural safety significance.
Based on a review of pertinent documentation, test results, and the concret_e placement packages, the TRT concludes that if nonconforming aggregate was used in the basemat of the Unit 1 reactor, it did not adversely affect its concrete properties. The only indication of water addition found bv the TRT was within,the stipulated limits, thus ensurir.g that there was nn adverse effect on the concrete. bowver, the absence of la'c oratory signatures on batch tickets repressts a" failure to follow QA/QC program requirements. The results of the; evaluation pertaining to the lack of laboratory signatures will be further' assessed as part of the overall progranmatic review concerning pr'oceduret,addresse,d under QA/QC Category 6 "QC Inspection." Therefore, the final acceptability 1of this evaluation yill be predicated on thedatisfactory results of the ' program-matic review of this subject. In its examination of 65 random samples of concrete placements in the turbine building, whic-h,is a nonsafety related structure, the TRT found no evide.nce of irregularities. Batch placements were within tolerances spp.cified by G&H, and the TRT found no docume,nta-tion that these slumpy equirements r had been violated. The placement of a single batch of concrete with a 4 -inch slump does not constitute.a violation of specificatiorts. The batch tickets state that none of the rejectedconcretebatcheswasplacedinthecirculstingwaterintake; therefore, the TRT conciddes that the allegation is without foundation. The alleger of AC-19 was not' identified so that the TRT could.not conduct a closing interview. The TRT is attempting to contact the individual who made allegation AC-21. The individual who made allegation AQC-16 did not wish to meet any further with the TRT;and will be informed of the pertinent TRT finding's by letter. The individual who made allegation AC-20 declined to be interviewed by the TRT and will also be informed of the pertinent TRT findings by letter. The alleger of AC-47 could not be located for a closing interview. 6.3 Actions Required: None. K-29 c j-. , - ,,
- 1. Allegation Category: Civil and Structural 2, Concrete Placements
- 2. Allegation Number: AC-22, AC-23 and AC-50
- 3. Characterization: It is alleged that " bad concrete work" and " sloppy" placement of concrete occurred at the Comanche Peak Steam Electric Station (CPSES) (AC-22, AC-23). It is also alleged that " soupy" concrete was placed in a slab in the Auxiliary Building in the summer of 1976 (AC-50).
- 4. Assessment of Safety Significance: The individual making allegations AC-22 and AC-23 was interviewed by the NRC Technical Review Team (TRT).
Allegation AC-50 was judged as having sufficient clarity for technical resolution without initial contact between the TRT and the alleger. 4 In testimony at an Atomic Safety and Licensing Board (ASLB) hearing, the first alleger did not identify a particular structure or concrete place-ment that had " bad concrete work" or that exhibited " sloppy" placement of concrete. To adequately encompass the concerns raised in this allegation, the TRT reviewed random samples of concrete placement packages from three safety-related buildings to determine if the allegations were valid. A review of 14 packages from the Auxiliary Building and 3 packages each from the Unit 1 Safeguards Building, the Unit 1 Containment Building (exterior), and the Unit 2 Containment Building (exterior) revealed that the quality control (QC) inspector accepted the forms and reinforcing steel placement prior to each concrete'placgment. Of the 23 placement packages reviewed, 10 had nonconformance reports (NCRs) related to con-crete placement. One package had four NCRs, two other packages had two NCRs each, and the remaining seven each had one NCR. Seven NCRs were resolved with the designation "use-as-is," seven with " repair," and one with " reject." The seven placements indicating " repair" were for concrete honeycombing; the one indicating " reject" was for the removal of concrete from a small pad. In addition to a records review, the TRT performed a walkthrough inspection of the safety-related buildings. The defects treated in these NCRs are visible from the surface and were examined by the TRT. The TRT condaded that there was no degradation in quality in any of the observable concrete surfaces. The TRT also interviewed two QC inspectors at Comanche Peak who were concrete placement inspectors on some of the concrete placements in the Auxiliary Building reviewed by the TRT. Both QC Inspectors stated that they were not cognizant of any " bad concrete work" and/or " sloppy" place-ment of concrete at CPSES. They stated that all personnel with construc-tion and concrete placement responsibilities would meet prior to each , placement to resolve any potential problems. They stated that for the l concrete placements they were involved with, the work was done in accord-ance with project procedures and other pertinent requirements. They also stated that placement crews cooperated with requests from QC personnel. The individual who made the allegations discussed above was contacted by the TRT to inform him of the TRT's finding. The alleger expressed his satisfaction with respect to the TRT's disposition of his allegations. j K-31
To investigate the allegation of " soupy" concrete in an Auxiliary Building slab, the TRT examined the following three placement packages, which included all the slab concrete placed during the summer of 1976: 002-7785-001, 002-2790-003, and 002-2790-004. The TRT noted that before l placing the first section, the contractor requested permission, which was granted, to place mortar rather than concrete in one area heavily con-gested with reinforcing bars. This might have been the " soupy" concrete s cited by the alleger. During placement of the three sections, five batches of concrete were rejected for excessive slump. In four of these cases, two or three cubic yards had been placed per the requirements of the ASTM l Standard Method for sampling fresh concrete (ASTM C 172). ASTM C 172 l requires that samples be taken at two or more regularly spaced intervals during discharge of the middle portion of the batch; and that samples not be taken from the very first or last portions of the batch. However, 70 to 80 percent of each batch was discarded. The concrete already in place was left in the forms. This type of occurrence is considered a normal procedure in concrete placement work and is judged to have no effect on safety.
- 5. Conclusion and Staff Positions: The TRT evaluated the allegations by reviewing a random sample of concrete placement record packages, by interviewing two former concrete placement inspectors, and by conducting a walkdown inspection of finished concrete work in three safety-related structures. This level of evaluation was deemed necessary to adequately encompass the potential scope of the allegations, which were not specific about where at Comanche Peak the " bad" and " sloppy" concrete work had been performed. In its records review, the TRT found some discrepancies in concrete placements that were identified and resolved by established QC procedures. However, the discrepancies found are not uncommon in concrete work; the TRT walkdown provided evidence that the discrepancies were resolved in that the concrete shows no degradation. The TRT also investi-gated the specific allegation concerning " soupy concrete" by reviewing all the relevant concrete placement packages and found the allegation to be without safety significance. The TRT found that mortar had been author-ized in lieu of concrete for a small portion of the structure. Accord-ingly, these allegations have neither safety significance nor generic implications.
The individual making Allegations AC-22 and AC-23 has indicated his satisfaction with the TRT disposition of his allegations. The alleger of AC-50 has not been located. The TRT is still trying to locate him for a closing interview.
- 6. Actions Required: None.
K-32
- 1. Allegation Category: Civil and Structural 3, Poor Weather Conditions
- 2. Allegation Number: AC-24, AC-35 and AC-52
- 3. Characterization: It is alleged that the placement of some concrete took place under the following adverse weather conditions: (a) during a rain-storm and without the approval of quality control (QC) personnel (AC-24) and (b) during or immediately before freezing weather (AC-35). It is further alleged that (c) there are several examples of field-cured cylinders which failed specification requirements, that some standard-cured cylinders failed specification requirements, and that the Schmidt rebound hammer test was misapplied in resolving problems created by these deficiencies (AC-52).
- 4. Assessment of Safety Significance: Allegation AC-24 was the subject of testimony given by Region IV inspectors, but the identity of the alleger, was not revealed. The TRT attempted to determine the alleger's identity but could find no record of it. Allegation AC-35 was judged as having sufficient clarity for technical resolution without initial contact betweer the NRC Technical Review Team (TRT) and the alleger. The TRT interviewed the alleger of AC-52.
- a. In assessing the allegation concerning placement during a rainstorm (AC-24), the TRT examined concrete placement package 101-8805-013 for a placement on the dome of the Unit 1 Containment Building. This package indicated that the final batch of concrete placed on the evening of January 18, 1979, was batched at 5:59 p.m. ; that only about 300 of the required 450 cubic yards had been placed; that the crew abandoned the placement in a heavy rain at 7:30 or 8:00 p.m. , leaving a gap with a 30-foot radius in the middle of the placement; that concrete batching started again at 8:00 a.m. on January 19; and that the lift was topped out at 12:21 p.m. There is no account of any irregularity during the shutdown. However, the craft personnel general foreman for the placement reported that the attempt to cover the partially completed concrete with plastic to protect it from the rain was not completely successful; that about a half cubic yard of concrete was washed out before the crew got the situation under control; that at about 10 p.m. he went to the batch plant, which was now empty because of the hour, dry-batched a half-cubic yard to the correct proportions, mixed it in two batches in the concrete labora-tory mixer, and placed it on the dome. At this ti.ne, all quality control personnel had gone home and were not available to approve or oversee the operation. This sequence of events was not refuted by the NRC Region IV investigation (inspection report 79-11) of this incident and is apparently correct. The TRT interviewed the author of the Region IV inspection report.
The action constitutes a violation of 10 CFR 50, Appendix B, Criterion X and indicates a partial breakdown of the quality control system. Following the completion of the dome, and after learning of the allegation of a violation, Brown & Root engaged Muenow and ' Associates to make an ultrasonic investigation of the portions of the K-33
dome potentially affected by the rainstorm. They also engaged Erlin, Hime, and Associates to interpret the Muenow report. The incident and the investigation are discussed extensively in nonconformance report (NCR) C-1418. The TRT reviewed NCR C-1418, " Final Report on the Concrete Evaluation in Dome Roof Section of Comanche Peak Unit 1," by Richard Muenow of Muenow and Associates, and " Discussion of Final Muenow Associates Report, Comanche Peak Steam Electric Station, Reactor #1 Dome Concrete Testing for Texas Utilities Service, Inc. ," by B. Erlin of Erlin, Hime, and Associates. The investigation revealed some minor discontinuities at 4 to 5 inches from the surface, and at 10 to 12 inches from the surface, with a few voids at a maximum of less than inch. Correlation of pulse velocity data on the dome with test cylinders containing the same materials indicated compressive strength in excess of 4,000 psi; this high pulse velocity, combined with the relative absence of voids, indicated a density in excess of 140 pounds per cubic foot. Accordingly, the structure as built satisfies the design requirements in the Final Safety Analysis Report. More convincing evidence of the acceptability of the dome concrete was provided by TUGCO's " Final Report on Structural Integrity Test for Unit 1 Concrete Containment Structure," CPDA-31, 792. The containment structure met all criteria for displacement and cracking control as well as structural rebound when subjected to 115 percent of design pressure.
- b. The allegation concerning concrete placed in freezing weather (AC-35) was in connection with the Safeguard Building basemat and the lowest level floor of the Unit 1 containment structure. The TRT reviewed in detail the relevant concrete placement packages, namely 105-2773-001 and 101-2808-001.
(1) Placement package 105-2773-001 reports on the Safeguard Building basemat, which was placed on December 31, 1975. All surface temperatures in the records comply with the specifications. However, it is alleged that on the seventh day of curing, when the ambient temperature dropped to 18 F, a portion of the con-crete in place was not protected by insulation. Brown & Root interoffice memo IM 4152 stated that all concrete was well covered with insulation except the edges, where it was difficult to place insulation because of protruding dowels, but that a careful examination of the concrete showed no evidence of damage caused by freezing.- Of the 15 field-cured specimens tested at 28 days, 12 failed the criterion of equalling or exceeding 0.85 of the laboratory-cured specimens. Of these, two failed the alternate criterion of exceeding the design strength by 500 psi. However, all results exceeded the design strength of 4,000 psi. The fact that 2 out of 12 failed to meet specification require-ments is not serious for concrete such as this, which was not loaded at an early age. The results of field-cured cylinder tests indicated that tne cold weather slowed the strength gain, but that the protection was adequate to attain the design strength in 28 days. Subsequent warmer temperatures provided all the strength required by the specifications. To compare the concrete near the dowels with concrete whose protection was not in doubt, the R.W. Hunt Co. ran.Schmidt hammer' tests on both the K-34
suspect concrete and the acceptable concrete at an age of 4 months. The results are recorded in HCP reports 10664 and 10849, which were inspected by the TRT. For both series of tests rebound num-bers ranged from 39 to 46. The concrete on the edge adjacent to the dowels, which was difficult to protect, is acceptable for the following reasons: (1) it was not exposed to freezing tempera-tures for 6 days following its placement; (2) concrete at that age should not be damaged by freezing; and, (3) Schmidt hammer readings were the same on suspect concrete as on well protected concrete. (2) Placement package 101-2808-001 reports on the concrete in the Unit 1 containment structure, which was placed on December 30, 1976. On the evening following the placement, the ambient temperature dropped below 20*F. The records showed a concrete surface temperature as low as 42*F during the first day and no surface temperatures below 50*F on subsequent days, in spite of the fact that ambient temperatures as low as 12 F were measured. The protection, as indicated by the records, complied with specifications. However, the allegation was triggered by an event detailed in Brown & Root (B&R) interoffice memo IM 7700. During the first evening, a TUEC QC inspector measured a surface temperature of 21*F. The B&R QC inspector noted that the TUEC inspector used an uncalibrated thermometer with a large range
'and took the reading in such a manner that the thermometer was not protected from the air so that in the B&R QC inspector's opinion, the TUEC inspector was measuring ambient temperature instead of the concrete surface temperature. Although the two discussed the adequacy of the technique, and a picture was taken of the technique, the records did not indicate that the matter was ever resolved.
To evaluate the condition of concrete alleged to have been exposed to freezing temperatures, the R.W. Hunt Co. ran Schmidt hammer tests on the suspect concrete and on concrete whose integrity was not in doubt. The results are in HCP report 22014, which was examined by the TRT. Rebound numbers for-suspect areas ranged from 25 to 35, and in sound areas from 27 to 36. 'The differences are not significant.
- c. The allegation concerning field-cured test cylinders, standard-cured test cylinders, and Schmidt rebound hammer tests (AC-52), is contained in the attachments to a letter, dated September 20, 1984, to Thomas Ippolito, NRC, from Mrs. Juanita Ellis, President of the.
Citizens Association for Sound Energy (CASE). The allegation states,
" Based on a review of documents attached and already in the record, it is apparent that the quality and compressive strength of the concrete at Comanche Peak is indeterminate at best, and, in some cases appears to be deficient." This observation'is presumably supported by Attachment D' to the letter, which lists 36 test cylinders in 18 placements with laboratory-cured strengths below 4000 psi. The alleger was interviewed, and ha stated that he was under the impression that the concrete was designed for a strength of 4000 psi. The TRT reviewed the records and'found that all the cited K-35
concrete was of designations C-301, C-302, C-305, or C-306, all of which have a design strength of 2500 psi. The lowest reported strength was 3267 psi. Thus, all these strengths met the strength specification by a wide margin. Furthermore, the TRT has discovered no standard-cured test cylinders in safety-related structures which failed the strength requirement. The allegation that some field-cured test cylinders failed to meet specification requirements is correct. The project specifications, 2323-SS-9, paragraph 7.3, cited the requireinents of ACI-318, the American Concrete Institute Building Code. These requirements are very conservative and are intended for building construction where slender flexural members are required to sustain a large portion of their design load at an early age. The requirement is that cold-weather protection shall be improved when the 28-day strength of field-cured cylinders is less than 85% of the strength of laboratory-cured cylinders. This requirement is more restrictive than is necessary for a massive structure. The definitive American Concrete Institute guidance on cold-weather protection is provided in ACI 306R-78, " Cold Weather Concreting." That document states that items such as foundations, substructures, and massive sections not subject to early load, which will be subjected to favorable curing temperatures prior to receiving the full design load, should be protected for 2 days if they are not subject to freezing in service and 3 days if they are subject to freezing. Protection is defined as maintenance of a temperature of 55'F for sections thinner than 12 inches, 50*F for sections 12 to 36 inches thick, 45 F for sections 36 . to 72 inches thick, and 40 F for sections thicker than 72 inches. No strength requirements are stipulated. All the field-curing deficiencies cited in Attachment D to the CASE letter, with three exceptions, fall into this less stringent category. The three exceptions are a slab in the Auxiliary Building, in which the field-cured strength was 3891 psi, and two cylinders representing slabs in the Safeguards Building with strengths of 3407 and 3956 psi. In these cases, the design strength was 4000 psi. The first and third had strengths sufficiently close to 4000 psi to eliminate any concern for safety. The second was in a region tested by the Schmidt rebound hammer and found to be equal in quality to sections of concrete whose quality was not in doubt. Even though more lenient criteria could reasonably have been established for much of the concrete, cold-weather protection was generally quite good. The ACI criteria for massive structures can produce field-cured strengths as low as 50% of the design strength at 28 days if the concrete is maintained at 35 F after protection is terminated. In contrast, most test cylinders at Comanche Peak exceed the design strength. Of 108 cylinders with a design strength of 4000 psi cited in Attachment D as failing to comply, 94% exceeded 3000 psi and the lowest strength was 2477 psi. Of 17 field-cured test cylinders failing to meet the design strength of 2500 psi, 14 exceeded 2000 , psi, and the lowest strength was 1820 psi. It also may be noted that ! field-cured cylinders usually underestimate the strength of the in place concrete they represent because they are not as massive and, therefore, benefit less from heat produced by hydration of cement during the curing process. K-36
The allegation questioned the use of the Schmidt rebound hammer for qualifying sections of concrete in which field-cured test cylinders failed to meet specifications. The above discussion makes the issue moot except for the single cylinder in a slab of the Safeguards Building. The use of the Schmidt rebound hammer has general acceptability and is specifically permitted by the Comanche Peak construction specifications as an aid in evaluating concrete strength in place, as discussed below. ASTM C-805 states, "The rebound number determined by this method may be used to assess the uniformity of concrete in situ, to delineate zones or regions of poor quality...." Paragraph 7.3.e of the project specifications states, " Evaluation of test results shall be in accordance with Section 17.1, 17.2, and 17.3 of ACI 301." Section 17.3.1 of ACI 301 states, " Testing by impact (Schmidt) hammer, soniscope, or other nondestructive device may be permitted by the architect / engineer to determine relative strengths at various locations in the structure as an aid in evaluating con-crete strength.in place or for selecting areas to be cored. Such tests, unless properly calibrated and correlated with other test data, shall not be used as a basis for acceptance or rejection." i Hammer results are normally not permitted as a substitute for laboratory-cured test cylinders, which form the basis for acceptance of the concrete. They may be used to judge the ade-quacy of protection or to determine when a portion of a structure may. be safely loaded. The ACI Building Code and the Comanche Peak specifications do not provide for the rejection of concrete on the basis of low-strength, field-cured cylinders. They merely require that protection be improved and that critical elements be cured for a longer period of time before being loaded. With the exception noted above, the low field-cured strengths were not in critical elements. The statement in Attachment D that all retesting which had been promised had not been carried out is a quality assurance matter, not a safety problem, and it will be investigated by the TRT QA/QC group. The statement that Schmidt hammer tests were not conducted on sections of concrete when both field-cured and laboratory-cured cylinders were below 4000 psi does not appear to be pertinent since there were no sections cited where both field and laboratory results were below the design strength.
- 5. Conclusion and Staff Positions: Although these allegations are true, they do not have structural safety significance.
(a) The Unit 1 dome was proved sound both by ultrasonic testing and by structural integrity testing. (b) Sections of concrete alleged to have been exposed to freezing temperature at an early age were shown by in place strength tests to have substantially the same strength as concrete whose protection was not in doubt. (c) The field-cured test cylinders demonstrated adequate protection for the type of concrete placed, with the exception of one slab in the Safeguards Building, which was shown by Schmidt hammer testing to be j adequate. K-37
L 4 Accordingly, these allegations have no structural safety significance. However, the results of the evaluation pertaining to the placement of concrete without QA/QC involvement, the response for improving protection when field-cured cylinders showed inadequate strer.gth, and the failure to carry out promised retests will be further assessed as part of the overall programmatic review concerning procedures addressed under QA/QC Category 6, "QG Inspection." Therefore, the final acceptability of this evaluation will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a sup)lementtothisSSER. 4 The TRT was unable to establish the identity of the individual who made allegation AC-24. The TRT cannot locate the alleger of AC-35 and has closed the allegation. The TRT has previously interviewed the alleger of AC-52, and a closure interview with the alleger is scheduled.
- 6. Actions Required: None.
N 1 I a l f K-38 6
l
- 1. Allegation Category: Civil and Structural 4, Concrete Voids and Cracking
- 2. Allegation Number: AC-25, AC-28, AC-32, AC-33 and AC-34
- 3. Characterization: It is alleged that the following concrete deficiencies occurred at Comanche Peak Steam Electric Station (CPSES):
- a. Hollow places existed in concrete behind the stainless steel liner of the Unit 1 reactor cavity (AC-25).
- b. Fresh concrete was placed on top of crumbling concrete during the construction of the spillway (AC-28).
- c. The repair of a 20-foot x 20-foot honeycombed area located in the Unit 1 Auxiliary Building was inadequate (AC-32).
- d. Cracks existed in the concrete reactor cavity wall of Unit 1 and in floor slabs in the plant buildings (AC-33).
- e. There are numerous concrete voids in building walls that can be located by tapping the walls with a hammer and listening for a hollow sound (AC-34).
Allegation AC-25 was investigated by Region IV and documented in inspection reports 80-08 and 80-11, which were reviewed by the TRT as a step in its own assessment of the allegation. In addition to these allegations, the Region IV resident inspector requested that the TRT review the following possible reportable design deficiencies involving concrete placing problems.
- f. Reportable Design Deficiency Concerns:
(1) A void was identified in the Unit 1 Reactor Building Steam Generator Compartment Wall. (2) On concrete placement 002-7810-002 at the 810-foot elevation of the Unit 2 Auxiliary Building, embedded foreign material was located with a flex drill.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) interviewed the individuals who made allegations AC-25 and AC-28. Allega-tions AC-32, AC-33, and AC-34 were made by former Brown & Root employees.
The TRT attempted to determine their identity but was unable to do so.
- a. The alleger originally stipulated that the hollow places were located behind the stainless steel liner of Unit I reactor cavity, but when interviewed by the TRT, he stated that he meant Unit 2. In assessing the allegation, the TRT interviewed the TUEC chief st*1ctural engi-neer who stated that when forms were removed from one section of the Unit 2 reactor cavity structure, honeycombed areas were discovered on the side of the structure accessible to visual examination. Because of the concern that the honeycombing indicated inadequate concrete ~
consolidation in this section and because the possibility existed K-39
that there might also be voids on the opposite side of the reactor cavity wall which were not accessible to visual examination, TUEC examined that section of the concrete wall ultrasonically. The examination revealed the existence of voids behind the stainless steel liner. Their existence and the required repair procedures are documented in Design Change Authorization (DCA) No. 6663. Repairs were being performed by TUEC at the time of the TRT review.
- b. There are two spillways at the CPSES, one located near the safe shut-down impoundment (SSI), and the other located at the Squaw Creek Dam.
The allegation did not specify which was intended, but the SSI spill-way was eliminated from consideration because it was constructed after June 1978, while the period cited in the allegation was 1976 and 1977. The Squaw Creek Dam spillway was constructed from August 1976 to January 1977. The TRT review of placement documentation indicated that the Squaw Creek spillway was placed in a single " lift"; therefore, no new con-crete could have been placed on hardened or crumbled concrete. During the interview with the alleger, it became apparent to the TRT from the types of placements being described that he had a general concern about the adequacy of cold weather placement practices during construction of the Squaw Creek Dam and appurtenant structures. How-ever, he was unable to identify a specific spot where specifications were violated. The TRT examined documentation for cold weather pro-tection for several placements during its investigation of other CPSES allegations. Those examinations confirmed that cold weather protec-tion was adequate. Furthermore, the Final Safety Analysis Report, Section 3.2, " Classification of Structures, Components, and Systems," indicates that the Squaw Creek Dam is not a seismic Category I struc-ture. Its failure would not affect safety during a safe shutdown earthquake.
- c. The concrete honeycombing referred to in the allegation is documented in nonconformance report (NCR) C-1034. The architect-engineer's direction was to remove the honeycombed area down to sound concrete and then fill the void area with dry pack concrete or small size coarse aggregate concrete, all in accordance with a standard, engineer-approved, repair procedure for such work. The TRT reviewed the repair procedure used (QI-QP-11.0-5) and believes it is adequate to properly repair the affected area. The repair is documented in Region IV Inspection Report 50-445/79-26. The NRC Resident Reactor Inspector-(RRI) observed various phases of the repair work from August 1978 through January 1979, when the repair was finally com-pleted. The RRI noted that the work was being done in an acceptable manner and in accordance with the approved instructions.
The TRT inspected documentation pertaining to the honeycombed area in the Auxiliary Building for concrete placement 002-7852-007 and verified that the area had been repaired. The TRT review of this concrete placement package revealed no documentation discrepancies concerning the repair. K-40 l _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _
- d. The existing cracks in the Unit 1 concrete reactor cavity wall have been the subject of a great deal of attention by the NRC and the designer. They have been documented in numerous NCRs, such as NCR C-650 and NCR C-1034. The TRT reviewed a random sample of the concrete placement packages for the Unit 1 Containment Building, Auxiliary Building, and both Safeguards Buildings, and found no evidence of specification violations during the concrete placement.
The TRT also inspected the cracks documented in NCRs 1034 and 650. The crack documented by NCR C-1034 is a small hairline crack, caused by shrinkage or thermal effects, that is so small that it cannot impair structural behavior and capacity. Cracks documented by NCR C-650 are evaluated in Civil and Structural Category 13.
- e. The NRC Resident Reactor Inspector (RRI) at Comanche Peak Steam Electric Station (CPSES) investigated this allegation (Region IV Inspection Report 50-445/80-16 and 50-446/80-16). The RRI learned that the alleger had worked at the site for 5 weeks in early 1980 in the Unit 1 Safeguards Building at the 790-foot elevation. The RRI found two locations at that elevation where a hollow sound could be obtained by tapping a wall with a hammer. He informed TUEC of this condition, and they found several more locations in the same general vicinity, all at the 790-foot elevation. Each area was marked and excavated to approximately 2 inches, that is, to the depth behind the first layer of reinforcing steel. The RRI observed several excavations and saw nothing abnormal about the concrete. He also queried the craft personnel who were excavating when he was not present and was informed that all excavations revealed nothing except uniformly solid concrete. The RRI tapped the concrete after it had been excavated to a depth of approximately 4 inches and could no longer detect a hollow sound. The allegation apparently was based on the premise that what the alleger interpreted as a hollow sound indicated a void in the wall. Excavations of the areas in question revealed no voids in the concrete.
- f. (1) Thi:; item was not the subject of an allegation. The TRT reviewed its disposition because it involved an issue similar to those raised in other CPSES allegations.
Nonconformance report (NCR) C-82-00858, which was reviewed by the TRT, indicates that a void did exist in the generator com-partment wall of the Unit 1 Reactor Building. As part of the NCR resolution, the matter was reported to Gibbs & Hill (Office Memorandum CPPA-21495, July 20, 1982) and they concluded that the wall would perform both its structural and radiation shield-ing functions whether or not the void was filled. However, to ensure that no safety issue could be raised, Brown & Root filled the void with nonshrink grout in August 1982, as documented in Inspection Report IR-C-6682. The TRT agrees that in its repaired state the wall presents no safety problem. (2) This item was not the subject of an allegation. The TRT l reviewed its disposition because it involved an issue similar to ' those raised in other CPSES allegations. K-41 1
The deficiency was documented in NCR C-82-01432, which was reviewed by the TRT. The TRT learned that a worker drilling holes for anchor bolts in a floor of the Auxiliary Building encountered an apparent void and debris. The debris appeared to be plywood chips. A Brown & Root examination of the area revealed that the drill had hit an embedded drain pipe and had removed'some of the foam insulation wrapped around the pipe per drawing MI-781. The driller had apparently mistakenly identified the foam as plywood. The disturbed insulation and concrete were then replaced, as documented in Brown & Root Inspection Report IR-C-7035. The TRT reviewed the Inspection Report and determined that the area was repaired in an acceptable manner.
- 5. Conclusion and Staff Positions:
- a. The allegation of hollow places in concrete behind the stainless steel liner of the Unit 2 Reactor Cavity is true and cannot be closed at this time. The area is currently undergoing repairs; the repairs must be inspected and approved by the NRC Resident Inspector before the TRT can determine that this issue has been adequately resolved.
The following allegations and concerns were found to have no structural safety significance.
- b. The TRT reviewed documentation for several placements done in cold weather and concludes that the protection was adequate. In addition, the allegation has no safety significance, since the spillway in question is not safety related.
- c. The allegation of honeycombing in the Unit 1 Auxiliary Building is true and the repairs made were in accordance with approved procedures. Therefore, the allegation has no structural safety significance.
There are numerous NCRs dealing with honeycombed concrete. Their evaluation and. subsequent concrete repairs are well-documented and did not result in allegations of improper construction except for those discussed herein. The quality assurance system apparently was adequate in documenting these repairs. However, there appears to have been a breakdown of quality control overseeing the consolidation of concrete as evidenced by the numerous NCRs and allegations AC-25 and AC-32. The results of the evaluation pertaining to inadequate consolidation of concrete will be further assessed as part of the overall programmatic review concerning procedures addressed under QA/QC Category 6 "QC Inspection." Therefore, the final acceptability of this evaluation will be predicated on the satisfactory results of t the programmatic review of this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER. i K-42
- d. While the allegation of cracking in the concrete basemat is accurate, it is not correct to assume that detrimental structural consequences will result from the cracks. The structures are designed to tolerate cracks of the magnitude and location of those found.
- e. The allegation of numerous concrete voids was not substantiated.
- f. (1) The reported void in the generator compartment wall of the Unit 1 Reactor Building is true. The void was filled even though it did not require filling from the standpoint of adequacy of design. The TRT determined that the wall in its repaired condition is safe.
(2) The area reported as containing unusual material in the concrete was adequately repaired so that this condition will have no impact on safety. The TRT will inform the individual who made allegation AC-25 of the TRT's findings by letter. The alleger of AC-28 has been notified by letter of the TRT disposition of his allegation. The allegers of AC-32, AC-33 and AC-34 are former B&R employees. The TRT was unable to establish their identity.
- 6. Actions Required: The repairs and the repair documentation to the honeycombing discussed in Item a must be inspected / reviewed and approved by the NRC Resident Inspector before the TRT can determine whether this issue has been adequately resolved. The successful completion of the repairs shall be reported to the TRT and will be verified by the NRC Resident Inspector prior to low power operations.
1 K-43
r i l
- 1. Allegation Category: Civil and Structural 5, Miscellaneous Concrete
- 2. Allegation Number: AC-26, AC-31, AC-36 and AC-43
- 3. Characterization: It is alleged.that the following irregularities occurred in connection with concrete construction:
- a. Equipment was set on grout before the grout properly gained strength through aging (AC-26).
- b. Hanger inserts were installed at improper angles (AC-31).
! c. Trash in the bottom of a form was covered with concrete (AC-36). AC-43 did not include any new allegations; it merely reiterated those made in AC-26, 31, and 36.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) interviewed the alleger of AC-31. Allegation AC-26 was judged as having sufficient clarity for technical resolution without initial contact between the TRT and the alleger. Allegation AC-36 was the subject of testimony given by Region IV inspectors, but the identity of the alleger was not revealed. The TRT attempted to determine the alleger's identity but could find no record of it,
- a. Allegation AC-26 concerns grouting of steel plates which were baseplates for the frames used to support parts of the internal assembly in Unit No. 2 when they were removed from the reactor pressure vessel. If the grout were damaged by the steel plate being loaded prematurely, the damage would occur immediately, while the grout was weak. If the grout survived the loading operation without damage, it probably would not suffer damage in use, since it gains strength rapidly while it is fresh and at a decreasing rate as it ages.
All elements of the internal assembly were located at the 860-foot. elevation. The TRT inspected all the grouted plates at the 860- and 862-foot elevations and found no evidence of grout failure. While the allegation may be true, all the grout survived the initial loading without damage. If this allegation is true, a quality control issue exists. The TRT Civil and Structural Group did not look into the QA/QC aspects of this allegation.
- b. It is alleged in AC-31 that Richmond anchor bolt inserts were installed between the 860- and 905-foot elevations in Unit 1 at angles not perpendicular to the concrete surface and that this condition was compensated for by use of tapered washers. The 4
allegation referred to discrepancies as great as ten degrees. The allegation was addressed in NRC Inspection Report 50-445/83-27, which was reviewed by the TRT. The TRT found that Brown & Root Procedure CP-CPM 9.10. " Fabrication of ASME-Related Component Supports," stated in Section 3.3.2 that: K-45
Surfaces of bolted parts in contact with the bolt or nut shall have a slope of no more than 1:20 with respect to a plane normal to the bolt axis. Where the surface of a high strength bolted part has a slope of more than 1:20, a. beveled washer shall be used to compensate for the lack of parallelism. Thus, inserts may depart 3 degrees from perpendicularity without any compensation and may depart further than 3 degrees if beveled washers are used. The procedure mentioned no upper limit on lack of perpendicularity. It did, however, stipulate that the predrilled holes in the tubular steel hanger safety-related supports may not be enlarged without prior approval. The TRT inspected 150 anchors between the 860- and 905-foot eleva-tions. The inspection consisted of a visual check of perpendicularity of the "as-installed" anchors, the occurrence of non perpendicular inserts without the compensation of using beveled washers, the maximum extent of insert deviation from perpendicularity, and the evidence of hole enlargements. Two were found to deviate from perpendicularity by more than 1:20; in these cases beveled washers were used. No hole enlargements were found. Thus, the TRT found no violation of the installation procedure. The allegation correctly asserts that some anchor inserts were not perpendicular to the concrete surface; how-ever, that in itself did not constitute a violation of procedure.
- c. Allegation AC-36 is concerned with trash from a Christmas party in December 1978, that was thrown into the form and was covered with concrete that was being placed on one of the two containment struc-tures. The alleged incident is extensively discussed in NRC Inspection Report (IR) 50-445/79-20, which was reviewed by the TRT.
Interviews with alleged participants, which were reported in IR 50-445/79-20, cast considerable doubt as to whether the party actually occurred. It was established in the inspection report that during December 1978 the alleger was at the project only on December 2, 3, and 4. The TRT obtained a printout of all concrete placements on the containment structures, and determined that the only placement which occurred during the period in question was on the dome of Unit 1 on December 3, 1978. The TRT examined concrete placement package 101-8805-002, which contained a complete narrative of the placement operation by the placement inspector. Notning unusual was noted, and both the formwork and cleanliness were checked as " satisfactory" on the checkout card. If anything unusual, such as dumping of trash, did take place, the structural integrity of the dome concrete was not compromised. The dome was proven to be adequate both in strength and in structural capacity, as indicated by the Unit 1 structural integrity test discussed in Civil and Structural Category 3. The TRT interviewed the individual who had raised the concern regarding the installation of anchor bolt inserts. This individual did not agree with certain TRT findings and provided the TRT with more information regarding his concern. The TRT investigated his concern further and scheduled another interview, but he declined to appear. K-46 l
- 5. Conclusion and Staff Positions: The TRT concludes that these allegations have no structural safety significance,
- a. All of the grout in question survived the initial load application without failure (AC-26). The possibility of premature loading will be assessed as part of the overall programmatic review concerning procedures addressed under QA/QC Category 6 "QC Inspection."
- b. No infraction of installation procedures for anchor inserts was found (AC-31).
- c. The allegation that trash was dumped into the bottom of a concrete form cannot be substantiated. Even if true, the containment structure concrete, including the dome, was shown to be adequate and acceptable in the in-situ structural integrity test (AC-36).
The TRT scheduled an interview with the alleger of AC-31 to discuss the TRT findings, but he declined to appear. A letter will be sent to him in lieu of a closing interview. The individual who made allegation AC-26 will be informed of the TRT's findings by letter. The TRT could not establish the identity of the individual who made allegation AC-36.
- 6. Actions Required: None.
O e K-47 i
- 1. Allegation Category: Civil and Structural 6, Rebar Improperly Installed or Omitted
- 2. Allegation Number: AC-30, AC-37, AQC-12, AC-38, AC-39 and AC-49
- 3. Characterization: It is alleged that reinforcing steel (rebar) was not properly inspected upon receipt at the site (AQC-12 and AC-37). It is also alleged that rebar was omitted in the following locations:
- a. A 6 foot x 6 foot section of concrete in the Safeguards Building (AC-30).
- b. The Unit I containment structure wall, specifically horizontal " tie" reinforcement (AC-38).
- c. Four column faces in the wall along column line EA of the Auxiliary Building (AC-39).
- d. It is also alleged that reinforcement was installed upside down in a building near the Unit 2 containment structure (AC-49).
In addition to these allegations, the Region IV resident inspector requested that the NRC Techni'.al Review Team (TRT) review the following possible reportable design r:eficiencies involving reinforcing steel (rebar): ,
- e. Reportable Design Deficiency Concerns:
(1) Rebar was omitted in a reactor cavity concrete placement between the 812-foot and 819-foot, 1/2-inch elevations in the Unit 1 Reactor Building. (2) Brown & Root construction requested a change in the configura-tion of two rows by nine layers of No. 9 reinforcing bars (2 x 9
- #9), as shown on drawing 2323-S1-0572, Rev. 4, to a continuous circular arrangement.
(3) Because of interferences with 14-inch diameter sleeves, the horizontal tails of No. 11 vertical reinforcing bars within the triangular columns surrounding the reactor cavity were modified to clear the sleeves. Also because of extreme congestion within the columns, stirrup details were modified. (4) Six No. 10 additional horizontal bars were omitted from a beam above a construction opening on column line KA between 6A and 7A in the Auxiliary Building. (5) Nine No. 9 and two No. 4 additional reinforcing dowels were omitted around an elevator shaft door in the Unit 1 Reactor Building. (6) Forty-six No. 9 dowels on the face of the wall in the excess letdown heat exchanger room in the Unit 1 containment structure were omitted. K-49
(7) Ten No. 8 additional horizontal dowels were omitted from a beam over a construction opening in Safeguards Building No. 1. (8) Brown & Root construction requested authorization to substitute No. 5 vertical wall rebars in lieu of the No. 8 wall rebars required in two corners of a wall in the Auxiliary Building.
- 4. Assessment of Safety Significance: The individual who made allegations concerning the improper receipt inspection of rebar (AQC-12 and AC-37) was not initially contacted by the TRT because the allegation was sufficiently clear to allow the TRT to proceed with its investigation. Allegations AC-38 and AC-39 concerning missing rebar in the Unit 1 containment wall and 4 columns in the Auxiliary Building were made by a former Brown & Root employee. The TRT attempted to determine this individual's identity, but could find no record of it. The TRT did not initially contact the individ-ual wh3 made the allegation concerning the rebar installed upside downThe (AC-49), because the alleger stated the problem had been corrected.
individual who made the allegation concerning the missing rebar in the Safeguards Building (AC-30) was contacted by the TRT to clarify his allegation. 1 The allegations that rebar was not properly inspected upon receipt (AQC-12 and AC-37) relate to the use of weldable reinforcing steel associated with the installation of radial At thisshear-bar location, reinforcement at the Grade 60,1-inch base steel x 4-inch of the containment structure. bars were joined by full penetration butt welds to No. 18 ASTM A615, Grade 60 reinforcing bars. Gibbs & Hill specification 2323-SS-10 required that a special chemical analysis be performed on each heat of reinforcing steel which was to be welded. Upon receipt, this reinforcing steel could be identified by the results of a special chemical analysis attached to the mill report. The QC inspector would verify that the results of the special chemical analysis conformed to the requirements of specification 2323-SS-10, and, if it was acceptable, QC personnel would then paint one end blue. It is alleged the No.18 Grade 60 reinforcing steel was used prior to the The TRT reviewed testimony proper inspection upon receipt by QC in 1975. taken during an interview in which the alleger stated that the QC inspector was pressured into hurrying the inspection process and that the reinforcing steel that was used prior to QC inspection was subsequently inspected and accepted by the QC inspector. This may indicate a partial breakdown in the area of QC receipt inspection of reinforcing steel. The TRT reviewed the receipt inspection reports for all No. 18 reinforcing bars received in 1975 and determined that three shipments were The received that had receipt a special chemical analysis attached to the mill report. inspection reports for these three shipments were signed off by QC, indi-However, the TRT coul.d not cating that an inspection had been performed. determine from its review of these receipt inspection reports whether any reinforcing steel was used prior to proper QC receipt inspection. The TRT's safety assessment for the remaining allegations and reportable design deficiencies are discussed below: l l K-50 l l l L _ -- >
- a. During an interview with the alleger, the TRT learned that the allegation of missing rebar in the Safeguards Building actually referred to the return pump station at Squaw Creek Dam (AC-30). For the detailed assessment of this allegation, see Civil and Structural Category 12, AC-29.
- b. This allegation (AC-38) was first reviewed in NRC Region IV Inspection Report No. 79-25, which refers to the omission of horizontal tie rebar in the Unit 1 containment structure, and j
concludes that the alleger was referring to an occurrence in the Unit 2 containment structure rather than in Unit 1. This event occurred shortly before the alleger terminated his employment, and it was assumed by the Region IV inspector to be the event to which he 1 referred. The omission of horizontal shear tie reinforcement in Unit
; 2 was originally investigated in Region IV inspection report 79-18, which notes that this reinforcement had been omitted near the junction of the containment wall and the hemispherical dome and was subsequently placed at a higher elevation. An analysis by Gibbs & ; Hill (G&H) concluded that the structure would be capable of carrying the design loads with the reinforcement in the as-built location. To determine if the allegation did indeed pertain to the Unit I containment structure and if all the reinforcing steel was placed in the Unit 1 containment wall as required, the TRT reviewed all 33 concrete pour packages (101-5805-001 through 101-5805-033) pertaining to the main concrete placements in the Unit 1 containment wall.
These pour packages contained rebar placement checklists which documented the results of inspections performed by B&R QC confirming the placement of the reinforcing bars to the applicable drawings. The TRT found three placement inspections in which the reinforcing bar placement was initially checked as unsatisfactory; the problems were then corrected, and the placement was signed off as satisfactory. The other 30 inspections performed were all checked as being satisfactory in that there were no deviations from the drawings.
- c. On October 27, 1977, a nonconformance report (NCR) C-806 was issued reporting the omission of 12 No. 8 vertical wall reinforcing bars at 4 column locations in the wall along column line EA of the Auxiliary
! Building (AC-39). The reinforcing steel had been omitted between the 3
810-foot, 6-inch and 831-foot elevations and involved four separate 1 concrete placements made from May to October 1977. This information was submitted to G&H engineering for resolution. G&H performed an j analysis which showed that the columns remained capable of carrying the design loads without the missing reinforcing bars and further directed that the bars be omitted from the columns for the remainder of their height through the 873-foot, 6-inch elevation.
- d. The TRT reviewed the April 10, 1979, transcript of a Region IV interview with an alleger and identified an allegation that reinforcement was installed upside down in a building near the Unit 2 containment structure (AC-49). However, during the interview the alleger claimed that the problem had been corrected prior to concrete placement.
l K-51
- e. (1) The reinforcing steel that slas placed between the 812-foot and 819-foot, -inch elevations in the reactor cavity wall of the Unit 1 Reactor Building was completed and inspected to draw-l ing 2323-51-0572, Rev. 2. After the concrete was placed, Brown
& Root received Rev. 3 to the drawing showing a substantial in-crease in reinforcing steel over that which was installed. G&H engineering was informed of the omission by Brown & Root non-conformance report C-669, which is referenced in the Brown &
Root internal deficiency report CP-77-6. G&H engineering replied that the omission of this additional reinforcing steel did not in any way impair the structural integrity of the structure. G&H stated that the additional rebar was added as a precaution against cracking which might occur in the vicinity of the neutron detector slots should a loss of coolant accident (LOCA) occur. A portion of the omitted reinforcing steel was placed in the next concrete lift above the 819-foot, 1/2-inch elevation. G&H stated that this was done to partially compensate for the rein-forcing steel omitted below and to minimize the overall area subject to possible cracking. The TRT requested documentation to indicate that an analysis was performed supporting this conclusion. The TRT was subsequently informed that an analysis had not been performed. l (2) In response to Brown & Root construction's Request for Informa-tion or Clarification (RFIC) RBCR-37, Design Change / Design Devia-tion Authorization (DC/DDA) No. 832 was issued stating that the configuration of the 2x9-No. 9 reinforcing bars (two rows by nine layers), as shown on drawing 2323-S1-0572, Rev. 4, could be changed to a continuous circumferential arrangement. The TRT reviewed this drawing and determined that these bars were among those omitted in the concrete placement between the 812-foot and t 819-foot, 1/2-inch elevations and subsequently placed above the 819-foot, 1/2-inch elevation (See e(1) above.) Revision 4 shows each of the four sets of No. 9 bars used to form the configura-tion required were to be bent in two places to form an approxi-mate circular configuration when placed. The DC/DDA stated the bars could be bent to a specified radius to form a true circular arrangement. The change, therefore, only affected the way in 1 which the bars were bent and did not reduce the load-carrying capacity of the structure. i (3) During the placement of reinforcing steel within the triangular j r columns surrounding the reactor cavity at the 826-foot, 11-inch I elevation, interferences were encountered. The horizontal tails of the No. 11 vertical reinforcing bars were interfering with 14-inch-diameter sleeves already in place. The TRT. reviewed i DC/DDA No. 6918 and the attached sketches which showed that six bars were cut and replaced with bars tailed up to achieve total anchorage and three bars were bent down to clear the sleeves. Also, due to congestion problems, the design of-the No. 4 stirrups surrounding the ten No. 18 circular bars was modified K-52
l to allow for installation. The stirrup design was modified to a two piece design rather than one piece, as originally designed. i This modification was permitted only within the triangular columns. (4) On October 26, 1977, nonconformance report (NCR) No. C-809 was issued by Brown & Root reporting the omission of six No. 10 additional horizontal reinforcing bars from a beam over a construction opening on column line KA between 7A and 6A in the Auxiliary Building at the 831-foot, 6-inch elevation. G&H engineering issued DC/00A No. 558 in response to the NCR. G&H engineering stated that the reinforcing bars were not required provided that one of the following conditions was met: (1) shoring remained within the construction opening until the slab above 831-foot, 6-inch elevation and the wall along column line KA above this elevation reached their design strengths, or (2) slab shoring remained adjacent to the construction opening until the concrete used to close the construction opening had reached its design strength. The intent was to provide adequate support to the 831-foot, 6-inch slab from either the wall above, the wall below, or from shoring. The disposition of the NCR showed that the shoring was left in the construction opening until the concr3te wall and slab above had cured. The TRT 1 reviewed the design change and solutions proposed and found the approach taken to be satisfactory. The TRT also reviewed drawing SA8-00711, which showed that the construction opening was closed with concrete pour No. 002-4810-042 on January 30, 1979. (5) Brown & Root issued NCR No. C-810 reporting the omission of nine No. 9 and two No. 4 additienal reinforcing dowels around the elevator shaft door in the Unit 1 Reactor Building at the 832-foot, 6-inch elevation. G&H DC/00A No. 477 indicated that the nine No. 9 dowels were to be drilled and grouted in place, and that the two No. 4 dowels could be placed without doweling into the slab. A review of the safety implications of the omitted reinforcing bars by Texas Utilities Electric Company (TUEC) Design Engineering showed that cracking of the concrete in this area could have occurred during conditions such as a seismic event if the reinforcing steel had not been placed. The review concluded that the cracking would not have affected the safety of the structure. (6) On October 31, 2.977, NCR C-811 was issued by Brown & Root reporting the omission of 46 No. 9 dowels on the face of the wall in the Excess Letdown Heat Exchanger Room in the Unit 1 Reactor Building. The civil QC inspector involved stated that the reinforcing steel had been installed and checked but that it was subsequently removed to allow for the installation of steam generator lower supports and reactor coolant pump tie supports and not replaced. G&H r.cgineering directed that the dowels be drilled and grouted in place. l K-53
{ (7) On October 21, 1977, concrete was placed which was to have contained ten No. 8 additional horizontal reinforcing dowels that were to run over the top of a construction opening in the Unit 1 Safeguards Building. NCR C-815 was issued by Brown & Root reporting this omission. In response to the NCR, G&H engineering decreased the size of the construction opening in the 7-S wall by placing a vertical construction joint 1 foot, 6 inches from the east face of the C-S/7-5 column. Decreasing the size of the opening allowed the ten No. 8 reinforcing bars to be placed with sufficient anchorage length developed by hooking the ends down into the 1-foot, 6-inch space. The TRT reviewed drawing S58-1065, which verified the decrease in opening size, and also showed the concrete pour numbers (105-4810-018 and 105-4810-034) for concrete placed in the 1-foot, 6-inch space and in the wall over the opening to the 829-foot, 6-inch elevation. A check of the rebar checklists included in these pour packages showed the rebar installation was inspected and accepted. Drawing SSB-1065 also showed that the construction opening was closed with concrete pour No. 105-4810-019. (8) Brown & Root construction issued request for information or clarification (RFIC) C-1987 on November 3, 1977, which requested authorization to substitute No. 5 vertical reinforcing bars in the wall 5 feet, 4 inches north of column line 3-A for the widths of the column line F-A and G-A walls (corner bars) in lieu of the No. 8 bars shown on the drawings. This involves the intersection of two walls. In assessing this issue, the TRT reviewed drawing 2323-5-0751, Rev. 15, which showed that the vertical bars in one of the walls, 5 feet, 4 inches north of column line 3-A between F-A and G-A, are No. 8 at 8 inches center to center (8 @ 8") each face and that the horizontal reinforcing is No. 6 @ 8" each face. Drawing 2323-5-0746, which shows the other walls involved along column lines F-A and G-A north of column line 3-A to be secondary walls. Drawing 2323-S-0785 gives the reinforcing requirements for secondary walls when the reinforcing is not otherwise noted on the elevation drawing. The walls along column lines F-A and G-A north of 3-A are 1 foot thick and require No. 5 @ 8" each way in each face. The No. 5 bars as installed in the walls along column lines F-A and G-A are, therefore, acceptable. In summary, one wall had No. 6 and No. 8 bars and the intersecting walls properly had No. 5 bars. The question involves the correct bars to use at the point of intersection (corners). Drawing 2323-5-0785 also indicates that where two walls intersect, the types of vertical corner bars used should be based on the thicker and/or more heavily reinforced wall. The four bars required in each corner are No. 8 based on the reinforcing in the wall 5 feet, 4 inches north of column line 3-A. The TRT reviewed DC/DDA 518, Rev. 1, dated November 9, 1977, which also verified that the No. 5 bars were acceptable for the wall but that the No. 8 vertical wall bars were to be installed in the corner as required. The TRT also reviewed concrete pour package 002-4831-017, which showed that the K-54 1
i l 1 l l l reinforcing steel installation as per DC/00A 518 Rev. I was inspected and accepted and that the concrete was placed on November 11, 1977. Therefore, the IRT concluded that the correct bars were used. The six documented structural sections with omitted reinforcing steel above indicate a breakdown in the quality control program as evidenced by the fact that these omissions were not detected prior to concrete placement.
- 5. Conclusion and Staff Positions: For the allegations concerning improperly inspected rebar (AQC-12, AC-37), the TRT concludes, based on the fact that the reinforcing steel used was subsequently accepted by QC, that this issue has no effect on the structural safety of the structure.
The TRT reached the following conclusions for the remaining allegations and reportable design deficiencies:
- a. Allegation AC-30, which refers to the return pump station at Squaw Creek Dam, not the Safeguards Building, is examined in Civil and Structural Category 12, AC-29.
- b. For AC-38, the TRT concludes that the horizontal shear bar reinforcement was placed in the Unit 1 Containment Building wall as required and further agrees with the conclusion drawn in Region IV Inspection Report No. 79-25 that the allegation refers to the Unit 2 containment structure, where the G&H analysis showed that the structure would be capable of carrying the design loading with the reinforcing steel in its as-built location. Therefore, the TRT concludes that this issue has no structural safety significance.
- c. The TRT reviewed the G&H analysis and agrees with their methodology and conclusion (AC-39). The TRT, therefore, concludes that this allegation has no structural safety significance,
- d. The TRT concludes that since this instance of improperly installed rebar was corrected prior to concrete placement, this issue has no adverse effect on the structural safety of the structure.
- e. (1) The TRT cannot determine the safety significance of this issue until an analysis is performed verifying that the reinforcing steel in the as-built condition is adequate.
(2) The TRT concludes that the change made to the No. 9 reinforcing bars did not affect the load-carrying capacity of the structure. (3) The TRT finds the modifications made to the interfering bars to be acceptable and to have no adverse effect on the structural safety of the structures. (4) The TRT finds that the omission of the additional reinforcing bars will have no adverse effect on the structural safety of the structure because shoring left in place until the concrete had cured made the additional reinforcing steel unnecessary. K-55
1 (6) The TRT concludes, based on the fact that the reinforcing steel was subsequently placed as per the disposition of the NCR, that there is no adverse effect on the structural safety of the structure. (6) The TRT concludes, based on the fact that the dowels were subsequently installed as per the disposition of the NCR, that this incident had no adverse effect on the structural safety of the structure. (7) The TRT concludes that by decreasing the size of the i construction opening, which allowed the reinforcing bars to be placed with sufficient anchorage length, this issue has no
' structural safety significance.
i (8) Based on the fact that the No. 8 vertical wall bars were installed in the corners as required, the TRT concludes this issue has no structural safety significance. However, the results of these evaluations which pertain to QC rebar placement and receipt inspection procedures will be further assessed as a
' part of the overall programmatic review concerning procedures addressed under QA/QC Category 6 "QC Inspection." Therefore, the final acceptability of these evaluations will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER.
Subsequent to its investigation, the TRT attempted to contact the indivi-duais who made the allegations discussed above to inform them of the TRT's findings. The individual who made Allegations AQC-12 and AC-37 will be iaformed of the TRT's findings by letter. The individual who made Allega-tion AC-30 was informed of the TRT's findings by letter. The TRT has not been able to contact the alleger of AC-49.
- 6. Actions Required: TUEC shall provide an analysis of the as-built condition of the Unit 1 reactor cavity that verifies the adequacy of the reinforcing steel between the 812-foot and 819-foot, -inch elevations. The analysis shall consider all required load combinations.
i K-56
- 1. Allegation Category: Civil and Structural 7, Uncontrolled Repair
- 2. Allegation Number: AC-10
- 3. Characterization: It is alleged that the removal of a Hilti bolt from the floor at the 852-foot level of the Safeguards Building resulted in a cone- W ped section of concrete being removed which was later repaired in an " uncontrolled manm r."
- 4. Assessment of Safety 5ignificance: The NRC Technical Review Team (TRT) did not initially attempt to contact the alleger because the allegation was sufficiently clear for the TRT to proceed with its investigation.
In assessing this allegation, the TRT examined NRC Investigation Report 81-12 (April 16, 1982), which described the observations of the area in question by an NRC investigator and the senior resident inspector. They concluded that the floor was repaired with a surface patch rather than being repaired all the way through. Such an uncontrolled and undocumented repair of a portion of a Category I structure indicates a lack of QA/QC control. The TRT concurred with these findings based on its observations of the floor area in question. Nevertheless, the TRT performed an independent evaluation of the safety significance of a 14-inch-diameter hole extending through the floor slab adjacent to pipe support No. CC-1-137-700-E63R, as alleged. This hole is located in the Electrical and Control Building and not in the Safeguards Building, as alleged, and as reported in NRC Investigation Report 81-12. For the worst-case analysis, the TRT assumed that two reinforcing steel bars (rebars) were cut in the process of removing the Hilti bolt. To account for the unknown quality of the material used in the repair, the TRT computed the ultimate moment capacity of the floor slab with and without a 14-inch section of slab removed. These estimated strength capacities were compared to the strength requirements necessary to resist the actual moments resulting from the slab design loads. The adequacy of shear capacity was also verified in a similar manner. From these analyses, it was evident to the TRT that the slab in its as-built condi-tion is capable of resisting the actual design loads, even though the most conservative engineering assumptions concerning cut rebar and a 14-inch hole were made.
- 5. Conclusion and Staff Position: Based on observations made by the TRT, the floor slab does not show any sign of degraded capacity or of poor repair i
practices. The slab appears continuous and composed of good materials. An independent TRT analysis of the slab capacity, based on conservative engineering assumptions, confirmed that the structural integrity of the slab would be maintained under its design loads. Accordingly, this alle-gation has no structural safety significance. K-57
However, the lack of QC inspection will be further assessed as a part of the programmatic review concerning procedures addressed under QA/QC Category 6, "QC Inspection." Therefore, the final acceptability of this uncor. trolled repair will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments to the existing con-clusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER. The individual who made the allegation was contacted by the TRT to inform him of the TRT's finding. The alleger expressed his satisfaction with respect to the TRT's disposition of his allegation.
- 6. Actions Required: None.
K-58
- 1. Allegation Category: Civil and Structural 8, False / Wrong Documents
- 2. Allegation Number: AQC-1, AQC-2, AQC-3, AQC-7, AQC-46 and AQC-51
- 3. Characterization: It is alleged, that the following records were falsified at various times:
- a. Concrete air entrainment records (AQC-1).
- b. Concrete laboratory test records (AQC-2). This allegation consisted of four separate parts: (1) that slump records were falsified, (2) that labo'ratory tests (air, slump, and temperature) for concrete placements of 10 cubic yards or less, prior to 1978, were not performed, (3) that laboratory tests were signed by a Level II inspector not present at the time the tests were performed, and (4) that the alleger signed a pressure gauge qualification test that he was not qualified to certify.
1 i c. Aggregate tests (January 1976). The alleger maintains that he and
, his foreman falsified these tests (AQC-3).
1 i d. Compression strength tests, at the direction of the general foreman and laboratory manager (AQC-7).
- e. Midpour tests during the placement of the Unit 1 Containment Building basemat on February 21, 1976 (AQC-46).
- f. Cadweld tensile test records were reported by an inspector without the tests actually being performed during the spring and summer of 1976 (AQC-51).
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) attempted to determine the identities of the individuals who made allega-tions AQC-1 and AQC-2 and could not find any record of their identities.
Allegation AQC-46 appeared in a April 1979, Fort Worth Star-Telegram article; the allegation was made by three unidentified individuals. The TRT did not initially attempt to contact the individual who made allega-tions AQC-3, AQC-7, and AQC-51 because the allegations were sufficiently clear to allow the TRT to proceed with its investigation.
- a. In assessing this allegation (AQC-1), the TRT reviewed documents contained in Brown & Root (B&R) Deficiency and Disposition Report 4
(DDR) No C-488 R1, R. W. Hunt Company QA Report HCP 21697 on concrete acceptance test results and the results of the Region IV investiga-tion of this allegation (inspection report 77-02). The records showed that on January 20, 1977, a 3.9 percent air content value was recorded in the concrete acceptance test report (HCP 21697) as 4.3 percent by a Level I inspector. The incident was reported to R.W. Hunt management by a co-worker. R. W. Hunt then issued a DDR identifying the placement of out-of-specification concrete and corrected the air entrainment value in the acceptance test reoort. The Level I inspector was subsequently fired for his action.
- K-59
t , To assess the possible safety significance of the falsification, the TRT examined the compressive strength test results for the concrete placement in question (105-7785-001) and found that the results ranged from 4905 psi through 5414 psi, and were well above the design specification strength of 4000 psi. The out-of-specification air content had little effect, if any, on the strength of the concrete placed.
- b. (1) The TRT cannot determine if slump test results were or were not falsified based on an examination of test records (AQC-2). To assess whether the records, if falsified, could have adversely affected the strength of the concrete, the TRT reviewed the results of compression tests performed on the concrete placed between April 11 and 13, 1978. (These dates correspond to the dates of the alleged falsification.) The TRT found the compressive strength to be consistent with that of concrete placed before and after the dates and within the specification.
(2) The allegation that laboratory tests (air, slump, and tempera-ture) were not performed on placements of concrete 10 cubic yards or smaller was investigated by the NRC Region IV staff (IE Inspection Report 78-07). This allegation was made in April 1978. The NRC Region IV staff reviewed log books that were the personal property of a number of laboratory personnel, but could not substantiate the allegation even though the alleger stated that such a review would be " revealing." The TRT reviewed all the concrete pour packages for the Pipe Tunnel, the Condensate Storage Tank and the Service Water Intake Structure, which are classified as safety related, to determine if any of the concrete placements were 10 cubic yards or less. The TRT identified eight concrete placements (111-1794-003, 111-1797-009, 111-1797-010, 111-1802-001, 111-9810-001, 035-9796-001, 035-9796-002, and 035-9796-003) that were placed prior to 1978 and that were 10 cubic yards or less. The dates of these placements were between August 1976 and February 1977. The eight concrete pour packages contained records showing field and laboratory tests results, but there was no way of determining whether the field tests were actually performed. However, for each of the above placements, concrete cylinders were also made and tested; the results demonstrated adequate strength. Prior to 1978, concrete placements in the Containment Structure, Fuel Handling Building, Auxiliary Building and Safeguards Building were generally for structural elements such as walls, slabs, and foundations. The placements 3 for these types of elements would generally be 50 yd or larger. To identify placements of 10 yd3 or less the TRT identified non-conformance reports concerning repair work (voids, honeycomb, etc.) to walls, slabs, etc. because placements for repair work would generally be less than 10 yd3 The TRT identified four concrete placements which needed concrete repair, but were repaired by means of " dry-pack" or " grout." The TRT interviewed four former R. W. Hunt employees who were involved in concrete testing activities at Comanche Peak during the time period in question. These four employees were employed on site with K-60
another employer. Three of them were working in the concrete testing laboratory. All four stated that they did not participate in or observe any falsification and/or failure to perform required concrete tests. (3) The allegation that a Level II inspector signed reports for tests performed on September 3 and 4, 1977, that he could have had no knowledge of was also reviewed by NRC Region IV personnel (IE Inspection Report 78-07). The alleger stated he had obtained this information from another individual who thought the falsification occurred in December 1977. The Region IV inspectors reviewed the daily payroll records of all laboratory personnel for the first 10 days of September and all of December 1977. The Level II inspector was present every day in September, but was absent Decembar 4, 5, 11 and 18 through 31. The Region IV staff could find no reports validated by the Level II inspector for the days alleged. The TRT reviewed the Region IV inspection report and concurred with the approach taken and the results of the investigation. In addition to reviewing the Region IV inspection report, the TRT examined strength test results for the concrete placed during the period stated in the allegation and found them to be above the minimum required design strength. (4) The allegation that the alleger signed a pressure certification test that he was not qualified to certify (on August 15, 1977) was investigated by NRC Region IV personnel (IE Inspection Report 78-07). Through an interview with Brown & Root calibration facility personnel, the NRC Region IV investigator learned that the pressure gauge was calibrated by Brown & Root personnel in accordance with their procedures. The calibration record was an R. W. Hunt form signed by the alleger who observed the test in accordance with the R. W. Hunt procedure. Prior to the Region IV investigation, B&R issued a DDR (February 17, 1978) that described this situation as a pre-existing and continuing problem in general, and proposed corrective action. The NRC Region IV staff concluded that while the allegation was substantiated, there were no safety consequences since the calibration was performed by a qualified individual in accordance with prescribed procedures. The TRT reviewed the Region IV inspection report, and concurred with the approach taken and the results of the investigation.
- c. The allegation (AQC-3) was first evaluated by the NRC Region IV staff (IE Inspection Report 79-09). The alleger, a former R. W. Hunt employee, stated that the falsification by him and his " foreman" occurred during the first 3 or 4 weeks of his employment, beginning January 19, 1976. The NRC Region IV staff reviewed the pre-qualification tests performed by Texas Industries, the aggregate supplier, on the material supplied to the site between January and May 1976 and also examined the results of in process concrete testing. Both sets of results complied with the specification requirements. The NRC Region IV staff also determined through discussions with a TUEC representative that the " foreman" was a Level K-61
II inspector in charge of the work. The NRC Region IV staff concluded that any falsification of test results on the part of the alleger would not have had a significant adverse impact on the quality of the concrete. The testing performed by Texas Industries was for the purpose of material qualification, whereas the tests performed by R. W. Hunt Company were to monitor the material for any deviation from the specification and to assure concrete of uniform workability and strength. The tests performed to verify concrete workability and strength were the test for slump and the cylinder test for compressive strength. The Texas Industries tests and the tests on fresh concrete indicated tha,t the aggregate was satisfactory for its intended purpose. In addition to reviewing the Region IV report, the TRT examined the results of slump and compressive strength tests for the period in which the falsification was alleged to have occurred. The test results were within specified limits and were consistent with concrete produced before and after this period.
- d. The 2 individuals making the allegation (AQC-7) and 13 other individuals were questioned by Region IV personnel between April 5, 1979 and May 7, 1979, regarding the allegation (IE Inspection Report 79-09). One of the allegers denied the allegation, stating he was misquoted in the newspaper. Another stated that he did not falsify concrete records himself but knew of other inspectors who had. One of the other 13 individuals interviewed stated he thought that falsification occurred, but did not know when or by whom. In addition, the NRC Region IV staff examined the test result statistics of the coacrete produced prior to and during the period of the alleged falsification and did not find any apparent variation in the '
uniformity of the concrete. The NRC Region IV staff concluded that the allegation could not be substantiated. The TRT staff reviewed slump and air entrainment test results of concrete placed during the period the remaining alleger was employed (January 1976 to February 1977) and did nct find any apparent variation in the uniformity of the parameters for fresh concrete placed during this period. However, since air content and slump tests have been alleged to be falsified, the TRT believes that additional action is required by TUEC to confirm that the results of the strength tests are representative of the strength of the concrete placed.
- e. According to an article that appeared in the Fort Worth Star-Telegram (April 1979), three unidentified R. W. Hunt Company concrete inspec-tors alleged that during the placement of 6600 cubic yards of concrete '
for the Unit 1 Containment Building basemat on February 21, 1976, some concrete was not tested, but instead the results were written in as averages (AQC-46). The concrete specification in force at this time required that slump, air content, temperature, and cylinders be taken every 100 cubic yards. The TRT reviewed concrete pour package (101-2805-001) for this placement and found 67 sets of test cylinders with the associated results of slump, air content, and temperature as per the specification. However, the TRT cannot determine from a review of these records whether the field tests were actually per-formed. Since the results of compression test's performed on the K-62
/
concrete cylinders would be the final measure of its acceptability, the TRT reviewed these results and found them to be acceptable and within the specification,
- f. The TRT reviewed all 440 Cadweld tensile test results for 1976 and identified 30 tests that were performed by the inspector in question (AQC-51). Twenty-eight of these tests were performed on one single day (October 13, 1976), while the other two tests were performed on two different days (July 21, 1976 and August 20, 1976). The TRT cannot determine whether or not all of the 30 tests in question were performed or if results were falsified and did not specifically look into the falsification issue. The remaining 410 tests performed by other inspectors all met the tensile strength requirements. The 30 Cadwelds tested were removed from the first layer of the exterior wall of the Unit 1 Containment Building at the 832-foot, 6-inch elevation. The TRT reviewed tensile test results of other Cadwelds performed by the individuals who made the Cadwelds in question. The results were found to be satisfactory. The Cadweld rejection rate for the 21 Cadwelders who made tne 440 Cadwelds ranged from zero percent to four percent, with one at six percent.
The fact that the allegations concerning the falsification of an air entrainment test and the certification of the pressure gauge test were substantiated indicates a partial breakdown in the QA/QC program in these areas.
- 5. Conclusions and Staff Position: The allegation (AQC-1) that a concrete air entrainment record was falsified is true. Even so, the compressive strength of the concrete in question was within specifications.
The allegation (AQC-2) that slump tests on April 11 and 13, 1978, were performed incorrectly and that the results were falsified could well be true and cannot be refuted. The TRT examined the compressive strength test results of the concrete in question nnd found that they were within specifications. The allegation (AQC-2) that laboratory tests for small placements were falsified was found to have no structural safety significance since, in addition to the recorded laboratory tests, the validity of which was questioned, cylinder strength tests were also performed to demonstrate adequate strength. In addition, in interviews with the TRT, former employees of the R. W. Hunt Co., who worked during the time period cited in the allegations, denied the validity of the allegation. Furthermore, the limited number of concrete placements of less than 10 cubic yards, even if improperly tested, would have little structural safety significance. The allegation (AQC-2) that an inspector signed test results for which he could have had no knowledge could not be substantiated because no reports could be found which had been signed by the inspector on the days alleged. Even if the allegation were true, test results showed the strength of the concrete placed during the period of the allegation to be above the minimum required strength. K-63
~ - . - - - . . - - . - - .- - - -- _ .- .- - . - ._ -
i i ! The allegation (AQC-2) that the alleger signed a pressure gauge test which
- he was not qualified to certify was found to have no structural safety j significance since the alleger did not actually perform the calibration.
l ' The TRT cannot determine the validity of the allegation (ACQ-3) that con-j crete aggregate tests were falsified. Nevertheless, the concrete placed during the period cited in the allegation was consistent with that of con- l crete placed before and after this period. The validity of the allegation (AQC-46) that'aidpour tests were falsified during the placement of the Unit 1 Containment Building basemat cannot be determined. The results of compression tests indicate that the concrete placed was of high quality. The TRT cannot determine the validity of the allegation (AQC-51) that - Cadweld tensile test results were falsified. If this f alsification did indeed occur, the strue N ral integrity of the exterior wall of the Unit 1 Containment Building has not been violated because (1) the tensile strength of other Cadweld test specimens performed by the 21 Cadwelders were found to be satisfactory, (2) the Cadweld rejection rate for each Cadwelder is at an acceptable level, (3) and the containment structure met all the criteria for displacement and cracking control as well as structural rebound when subjected to 115 percent of design pressure, as 1 stated in CPDA-31, 792, " Final Report on Structural Integrity Test for Unit 1 Concrete Containment Structure." Accordingly, the above allegations have no structural safety significance. However, the allegations resolved on the basis of acceptable concrete 4 strength test results may need to be further assessed pending the j resolution of allegation AQC-7. Also, the results of these evaluations ' pertaining to QC inspection procedures will.be further assessed as a part of the overall programmatic review concerning procedures. addressed under QA/QC Category 3, " Records." Therefore, the final acceptability of these evaluations will be predicated on the satisfactory results of the programmatic review an( the satisfactory resolution of allegation AQC-7. Any adjustments to the existing conclusions of these evaluations will be reported in a supplement to this SSER. ,
)
The allegation (AQC-7) that compressive strength test results were falsified cannot be closed at.this time. The TRT agrees with the Region-IV staff that the uniformity of the fresh concrete placed during this-period suggests that there was no serious problem with the hardened l
- concrete and, therefore, no serious safety problem. However,' this conclusion is based on air content, slump, and strength tests, all of
'L which have been alleged to be falsified. The issues regarding air content and slump, as well as other allegations discussed above, were resolved on i' the basis of the concrete strength test results. Due to the importance of the concrete strength test results, the TRT concludes that additional-action by TUEC is necessary to provide confirmatory evidence that the reported concrete strength test results are indeed representative of the strength of the concrete placed in the Category I structures. ' The TRT is attempting to contact the individual who made allegations AQC-3, AQC-7, and AQC-51 to inform him of the TRT's findings. 3 K-64 4 t
-.,... - .~ ,_r,- --,.,.- - ,, . , . ~ , - ,-- ...,_,y-, -rw, , . . ---_,,.m.. - , , m,-.. ,m .,-.,.n,_m,w e r- -, 4 -,-,--w-ee --vr , ..w.- - - ..-i.. , , - + .
- 6. Actions Required: TUEC shall determine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a program to assure acceptable concrete strength. The program shall include tests such as the use of random Schmidt hammer tests on the concrete in areas where safety is critical. The program shall include a comparison of the results with the results of tests performed on concrete of the same design strength in areas where the strength of the concrete is not questioned, to determine if any significant variance in strength occurs.
TUEC shall submit the program for performing these tests to the NRC for review and approval prior to performing the tests. l I K-65 i
- 1. Allegation Category: Civil and Structural 9, QC Inspector Training
- 2. Allegation Number: AQC-9
- 3. Characterization: It is alleged by two former R. W. Hunt Company employees that (a) after a March 1977, NRC investigation, closed book recertification tests of R. W. Hunt inspectors were done "open book" and that (b) tests were given with the answers provided.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) did not initially attempt to contact the two allegers because the TRT was able, with some initial investigation, to clarify the allegations suffi-ciently to proceed with its evaluation.
The allegation about the recertification tests refers to recertification testing that was required because a Region IV investigation in 1977 (inspection report 50-445/77-02) questioned the most recent certification of R. W. Hunt Level I and II inspectors. The NRC Region IV staff found that R. W. Hunt did not comply with the minimum 2 year experience requirement for qualification as a Level I concrete inspector, as required by the ASME Code to which they were committed by the Preliminary Safety Analysis Report (PSAR). The R. W. Hunt Skills Training Certification { manual stated that " Experience requirements may be reduced if the individual can demonstrate capability in a given job through previous performance or satisfactory completion of an examination and orientation training." Also, the Region IV staff found that the " Certification of Qualifications," which was issued to each Level I and II inspector did not include the activity the inspector was qualified to perform or the basis used for certification, as was required. Each candidate for certification was required to demonstrate proficiency in performing specific practical tests on one or more samples approved by the Level III examiner. Tne Region IV staff found that R. W. Hunt had permitted a Level I inspector to perform concrete cylinder compression tests and aggregate sieve analysis without evidence of demonstrated proficiency and approval in accordance with the above requirements. As a result of this investigation, Brown & Root (B&R) audited R. W. Hunt training and certification activities and required each inspector to be ; recertified by attending specific training sessions and by closed book l testing. The work performed by the personnel qualified under the previous provisions was reviewed by B&R QA personnel and found to be within the , specification requirements. In addition B&R assigned a QC civil engineer l to work full-time with R. W. Hunt on site to ensure full compliance with ; project requirements. j Between April 3, 1979, and May 7, 1979, NRC Region IV inspectors inter-viewed 15 individuals associated with concrete testing activities regard-ing the allegation concerning the recertification tests (inspection report 50-445/79-09), including the 2 who had originally made the newspaper allegations (April 1979). The alleger who stated in the newspaper article that after the NRC investigation of March 1977, the recertification of K-67 1 l l l __ , _ _ _ - .. - - - - - - -- --
1 inspectors to test cy'linders was "open-book" did not mention open-book testing when interviewed by the Region IV inspectors. He stated that he had failed a Level I soils test, that he was subsequently given answers (orally) by the laboratory manager, and then repeated the soils test using the notes he had taken. He also stated that to obtain his recertification (after March 1977) he needed only to have a supervisor sign the recertification form. When interviewed by Region IV, the other individual who alleged in the newspaper that he had been given answers to tests reaffirmed his allegation. To determine if the individual was referring to the recertification tests or to tests he had taken prior to March 1977 to obtain his initial certi-fication, the TRT reviewed his employment records. The TRT found that this individual was not employed by R. W. Hunt at the time the March 1977 recertification tests were given. His allegation therefore would be referring to tests he took prior to March 1977 to obtain his initial certification. Two other individuals who were questioned by the Region IV staff between April 3, 1979 and May 7, 1979 generally supported the allegations. Again, to determine which tests these individuals were referring to, the TRT reviewed their employment records and the results of the Region IV interviews. One individual states that the recertification tests were administered properly; the other was not employed by R. W. Hunt Co. at the time the recertification tests were given. Therefore, these two individuals would be referring to tests they had taken prior to March 1977, their initial certification tests. These three individual's certi-fications were among those questioned by the 1977 Region IV investigation (inspection report 77-02) and their work had been audited and found acceptable. In summary, the allegation that recertification tests were administered "open-book" was supported only by the individual making the allegation. The allegation that test answers were given for tests taken prior to March 1977 was supported by three individuals. Eleven other individuals who were questioned did not support the allegations. The TRT reviewed the personnel file of the individual who made the allega-tion concerning the open book recertification tests and learned that he was employed from August 16, 1976, to June 28, 1978. Therefore, he was among those inspectors whose previous work had been reviewed and found acceptable. The TRT also found copies of tests taken by the-individual for recertification in concrete and soil inspections. The TRT reviewed the test the individual had taken relating to concrete cylinder tests and could not determine whether the test had been administered properly. The TRT also examined test result statistics for concrete placed from 1975 to 1978, and found that the concrete placed was of uniform quality and strength and that there was no apparent variance in the test results. Approximately 35 different inspectors were involved in concrete testing between 1975 and 1978; more than 7 inspectors conducted concrete com-pression tests on a rotating basis during this period.
- 5. Conclusion and Staff Position: The allegation that answers to tests were given prior to March 1977 cannot be refuted. An NRC Region IV investi-gation (Inspection Report 77-02) questioned the qualifications of the R. W. Hunt inspectors. The work performed by the R. W. Hunt inspectors K-68
certified prior to March 1977 was reviewed by B&R and was found to be within specifications, a fact subsequently reported to NRC Region IV staff. Therefore, the TRT concludes that this allegation has no structural safety significance. The allegation that the recertification tests for concrete cylinder testing were given "open book" cannot be substantiated. This allegation was not supported by any of the other individuals questioned, which sug-gests it was an isolated occurrence. The work performed by this individual prior to March 1977 was audited and found to be satisfactory, which would indicate the individual possessed the knowledge required to properly per-form the required testing. In addition, the test results for concrete placed, including the concrete compression tests, were contributed to by many inspectors whose qualifications were acceptable. These test results showed the concrete was of uniform quality and strength. Based on the fact that the inspector's work had previously been reviewed and found to be acceptable, and that a number of inspectors contributed to the test results, which showed the concrete to be of uniform quality, the TRT concludes that this issue has no structural safety significance. The results of these evaluations will be further assessed as a part of the overall programmatic review concerning inspector qualifications addressed under QA/QC Category 4, " Training and Qualification of Personnel." There-fore, the final acceptability of these evaluations will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments'to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER. The TRT contacted one of the allegers who made one of the allegations discussed above. He declined to meet with the TRT. He will be informed by letter of the TRT's findings. The TRT is attempting to contact the other alleger involved.
- 6. Actions Required: None.
4 K-69
- 1. Allegation Category: Civil and Structural 10, Improper Testing
- 2. Allegation Number: AQC-4, AQC-5, AQC-6, AQC-8, AQC-11 and AQC-48
- 3. Characterization: It is alleged that the following violation of testing procedures occurred:
- a. Equipment required for aggregate testing was sitting unused on laboratory shelves (AQC-4).
- b. Shortcuts were taken on tests involving grading of aggregate (AQC-5).
- c. During the placing of a 6600-cubic yard section of the basemat for Unit 1, some concrete was placed without the required testing (AQC-6).
- d. Concrete cylinder compression tests were run at a faster loading rate than permitted by NRC regulations (AQC-8),
- e. Concrete test cylinders with adequate strength were used to represent other placements (AQC-11).
- f. Concrete test cylinders in the Hunt Laboratory moist room were allowed to dry (AQC-48).
Allegations AQC-4, AQC-5, AQC-6, and AQC-8 were investigated by NRC Region IV and documented in inspection report 79-09, which was reviewed by the NRC Technical Review Team (TRT) as a step in its own assessment of the allegations.
- 4. Assessment of Safety Significance: Allegations AQC-4, AQC-11, and AQC-48 were judged as having sufficient clarity for technical resolution without initial contact between the TRT and the alleger. Allegations AQC-5, AQC-6, and AQC-8 were made in newspaper articles which did not identify the allegers, and the TRT has been unable to determine the allegers' identities.
- a. The test equipment that allegedly remained unused at the project laboratory was for the test for Potential Reactivity of Aggregates (Chemical Method), American Society for Testing and Materials (ASTM)
C 289. Test Laboratory Manual TLM-004 (CP-QP-0.5), which was in effect during most of the construction period, required that the test be run once for each 4000 tons of aggregate. The TRT inspected
" Folder 1 - Potential Reactivity, 4000 Ton Test" and learned that between May 6,1975, and July 12, 1978, there were 60 tests for potential reactivity. The TRT also interviewed the laboratory tech-nician who performed most of the tests. This period covers the bulk of heavy construction and the entire employment period of the alleger.
The testing rate during this period exceeded one test per 4000 tons of coarse aggregate. Thus, testing was at a higher rate than required by the testing requirements. K-71
- b. The shortcut alleged is that TUEC used a hot plate for drying aggregate in its sieve analyses of coarse aggregate rather than an oven, as specified in test method ASTM C-136. Note 4 of that method contains the following information: -
Samples may be dried at the higher temperatures associated with the use of hot plates without affecting results, provided steam ) escapes without generating pressures sufficient to fracture the particles, and temperatures are not so great as to cause chemical breakdown of the aggregate. The alleged shortcut, then, is permitted by the provision just cited.
- c. The TRT inspected all batch tickets and test records for the 6600-cubic yard basemat placement and physically inspected those portions of the mat still accessible. A placement that size required 66 sets of test cylinders, with associated data on fresh concrete.
There were 67 sets of records in the file, all of which showed compliance with specifications. While little of the placement was available for inspection, that portion that could be seen was in excellent condition. The implied aspect of falsification is dealt with in Civil and Structural Category 8.
- d. Cylinder strength testing must be done in accordance with " Method for Compressive Strength of Cylindrical Concrete Specimens," ASTM C-39.
That method permits any rate of' loading during the first half of the loading range, but res ,ricts the rate of loading at fracture to the range of 20 to 50 psi per second. A higher rate of loading may produce a higher indicated strength. The definitive work on investigating the effect of rate of loading on indicated strength (Watstein, D., "Effect of Straining Rate on the Compressise Strength and Elastic Properties of Concrete," Proceedings, American Concrete Institute, Vol. 49, 1953, p. 729) demonstrated a sighificant increase in indicated strength for very high dynamic rates of loading. However, a rate 100 times that specified produces an indicated increase in strength of only 10 percent. The testing machine used to break cylinders on the Comanche Peak project, a Forney Model CAC-50-DR, if run at maximum capacity, could achieve a testing rate no greater than 20 times the specified rate. This rate could produce an apparent increase in strength of about 6.5 percent. For 4,000 psi concrete the apparent increase would be about 250 psi. In a detailed check by the TRT of several placement packages, and a spot check of others, the 4000 psi concrete averaged more than 5000 psi, and individual results exceeded 4500 psi. Thus, if some tests were conducted at too high a loading rate, no results were changed from failing to passing. If there were tests within 6.5 percent of the design strength not detected by the TRT, the strength could be expected to gain 6.5 percent within a few weeks, so that the design strength would be attained long before the structure was put into service.
- e. The TRT investigated the number of cylinders available for switching to other placements. The alleger stated that the switch occurred af ter "a good sample was found." By the time the 28-day tests were K-72
completed, at most, two extra cylinders remained for which the test results could be switched to the testing data for other placements. l Unless this was a widespread practice, its significance would be small because of the relatively few cylinders available. The allegation does, however, raise a question as to the effectiveness of quality control in the laboratory.
- f. To investigate the allegation that concrete cylinders in the laboratory moist room were allowed to dry, the TRT examined the current procedure for documenting moist room conditions and interviewed a Level II inspector who was present throughout the period when the laboratory was operated by the R. W. Hunt Company.
At present there is a thermometer which provides a permanent temperature record. While there is no quantitative measurement of humidity, there are daily visual observations of the presence or absence of fog in the room. These observations are also a part of the record. During the R. W. Hunt operation, temperatures were recorded, but there apparently was no documentation of humidity. Batch plant inspectors were required to note the condition of the moist room, but there is no record of their observations. There is a history of breakdowns in the water supply to the laboratory, and a shutdown as long as 6 hours has been documented. With the door to the moist room closed, there would be a negligible drop in humidity during such a period. As long as the relative humidity remains above 90 percent, concrete curing conditions are favorable. It is pertinent to note that any drying that might occur would produce conservative results in that measured strengths would be lower than actual strengths.
- 5. Conclusion and Staff Positions: The TRT concludes that these allegations have no impact on structural safety.
- a. All required tests for ASTM C-289 were performed.
- b. The alleged shortcut in carrying out aggregate grading tests is permitted by the provisions of the specified test method in ASTM C-136.
- c. All required testing was carried out in connection with the 6600-cubic yard basemat placement.
- d. Although this allegation may have been true, the fastest possible loading of test cylinders would have increased the indicated strengths by no more than 6.5 percent and would have had no effect on the acceptability of the concrete.
- e. The alleged substitution of test cylirders is unlikely to have affected a sufficient number of cylinders to have had a material effect on the overall test resultr..
- f. Although the allegation that the laboratory failed to maintain the water supply at all times may be true in that there were brief K-73
f shutoffs of water to the moist room humidifiers, these periodic breakdowns would result in conservative strength results on concrete cylinders. Accordingly, these allegations have no structural safety significance. However, the effectiveness of quality control in the laboratory will be further assessed as part of the overall programmatic review concerning procedures addressed under QA/QC Category 6, "QC Inspection." Therefore, the final acceptability of this evaluation will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER. The alleger for allegations AQC-4 and AQC-48 cannot be located. Allega-tions AQC-5, AQC-6, and AQC-8 were made in newspaper articles in which the allegers were not identified. The TRT has been unable to determine their identities. The TRT sent a letter to the alleger of AQC-11 explaining its disposition.
- 6. Actions Required: None.
K-74
- 1. Allegation Category: Civil and Structural 11, Seismic Design / Construction
- 2. Allegation Number: AC-41
- 3. Characterization: It is alleged that there was poor workmanship regarding the use of elastic joint filler material ("rotofoam") as a temporary spacer during construction to maintain the required air space between seismic Category I structures.
- 4. Assessment of Safety Significance: This allegation was received anony-mously; therefore, the TRT could not contact the alleger about its evalua-tion of AC-41.
TUEC informed NRC Region IV on November 23, 1977, of this allegation, which TUEC received anonymously in a telephone call on November 22, 1977. A Region IV inspector reviewed the allegation during an inspection con-ducted between November 28 and December 2, 1977, and concluded, based on the information available to him at the time, that all temporary rotofoam had been removed from the seismic gap between Category I structures. The matter was left open pending a Region IV review of the Brown & Root (B&R) QA/QC inspection and documentation program, which was being initiated to assure that the required seismic gap between Categcry I structures was being maintained. Rotofoam was used as a temporary spacer during cons-truction to maintain this gap. Once the concrete hardened, the rotofoam was ren.oved to eliminate any load transfer or dynamic interaction between
, buildings. If the relative motion bet' ween buildings was small and the presence of rotofoam was considered in the dynamic analysis of the build-ing, leaving the rotofoam in place may not have had a significant impact on the dynamic performance of the buildings.
During an inspection between January 3 and 13, 1978, the Region IV inspector reviewed B&R procedure CP-QCI-2.4-9, " Inspection of Elastic Joint Filler Material Removal," Revision 1 (December 12, 1977), and B&R inspection reports for December 15, 1977, and January 3, 1978, and had no further questions regarding this matter. The NRC Technical Review Team (TRT) attempted to obtain a further clarification of the concerns expressed by the alleger; however, neither TUEC nor the Region IV office had records of the alleger's telephone conversation other than what is stated above. The TRT determined, however, that prior to the time the allegation was made there was a misunderstanding as to whether or not the rotofoam should remain in place as part of the final construction. A letter from Gibbs & Hill (G&H) of September 6,1977 (GTT-1543), indicated that construction was improperly proceeding on the basis that the rotofoam could be left in place. The letter further stated that this assumption was not in accordance with the facility design drawings and design concept and that expansion joints above grade should consist of a clear gap between buildings, i.e., free of rotofoam. As noted in the G&H letter, it was intended that the rotofoam be left in place below grade. Since construction had proceeded above grade, TUEC instructed B&R, in a letter of October 7, 1977 (TUS-5012), to remove the rotofoam above grade. As noted, B&R procedure CP-QCI-2.4-9 was also implemented to verify removal of the rotofoam. Based on discussions l K-75 l
1 with TUEC and G&H engineers, the TRT found that the rotofoam was to be left in place for the expansion joints above grade between the Safeguards , Building and the Reactor Building. If properly implemented, B&R procedure CP-QCI-2.4-9 should have provided an adequate inspection record for demonstrating that the air gap between buildings was adequately maintained. However, the TRT found only two inspection reports relating to this procedure (the December 15, 1977 and January 3, 1978, reports referenced). These reports did not fulfill the Furthermore, this complete inspection requirements of CP-QCI-2.4-9. procedure was deleted on July 18, 1978 (B&R memo IM-14835). A G&H memo of January 30, 1978 (GHF-2390) indicated that an inspection was made on November 23, 1977, and stated that the removal of the rotofoam from the subject areas was acceptable. However, the memo related only to construc-tion at that point and did not provide any documented evidence of the inspections that were made. A B&R interoffice memo of February 19, 1978 (IM-12934), discussed an inspection of the seismic gap between the Auxiliary Building and the Containment Building for Unit 1. The memo indicated that the removal of rotofoam was not completed and requested further removal and/or engineering evaluation. TUEC engineers apparently investigated this matter; however, the TRT found no formal documentation indicating the resolution of this matter. Between September 14, 1978, and October 17, 1978, a B&R QC inspector made additional inspections of the air gap between seismic Category I structures. Six different areas were inspected. In five out of the six areas, the inspector indicated unsatisfactory conditions due to the presence of foreign material in the air gap, such as wood wedges, rocks, clumps of concrete, and rotofoam. These unsatisfactory inspection reports were officially resolved on April 18, 1983, in response to NCR C-83-01067 (April 13, 1983). The disposition of this NCR noted that " fieldBased investi-on gation reveals that most of the material has been removed." discussions with TUEC engineers, it is the TRT's understanding that field investigations were made but that no permanent records of these investi-gations were maintained. TUEC engineers provided the TRT with five pages of field measurements made between March 15 and March 24, 1983, which indicated that investigations of the air gap between the Auxiliary Building and the Fuel Building were conducted. These measurements appeared to indicate that the required air gap was not provided to the 813-foot, 6-inch elevation (the required elevation in procedure CP-QCI-2.4-9). Even though the measurements indicated a nonconforming condition, TUEC could not provide any documentation indicating whether an engineering analysis was performed to justify this nonconformance or whether the material was subsequently removed. The TRT attempted to inspect the air gap between the structures but could not because in most cases the final joint sealer or roof flashing had already been installed. In several areas between the Auxiliary Building and the Safeguards Building, the air gap could be observed and appeared to be clear of any obstructions. In one doorway between the Safeguards Building for Unit 1 i i K-76 l - -- , . .- ._ 7 -. . . ,-
i and the Auxiliary Building at the 830-foot, 6-inch elevation, the air gap was clear to an observer looking up. However, a wooden board and other debris were observed when viewed straight in and downward.
- 5. Conclusion and Staff Positions: Based on the review of available inspection reports and related documents, on field observations, and on discussions with TUEC engineers, the TRT cannot determine whether an adequate air gap has been provided between concrete structures. Field investigations by B&R QC inspectors indicated unsatisfactory conditions due to the presence of debris in the air gap, such as wood wedges, rocks, clumps of concrete and rotofoam. The disposition of the NCR relating to this matter states that the " field investigation reveals that most of the material has been removed." However, the TRT cannot determine from this report (NCR C-83-01067) the extent and location of the debris remaining between the structures.
Based on discussions with TUEC engineers, it is the TRT's understanding that field investigations were made but that no permanent records were maintained. In addition, it is not apparent that the permanent installation of elastic joint filler material ("rotofoam") between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the seismic analysis assumptions and dynamic models used to analyze the buildings, as these analyses are delineated in the Final Safety Analysis Report (FSAR). The TRT, therefore, concludes that TUEC has not adequately demonstrated compliance with FSAR Sections 3.8.1.1.1, 3.8.4.5.1, and 3.7.B.2.8, which require separation of seismic Category I buildings to r* event seismic interaction during an earthquake. Depending on the extent of nonconformance with FSAR Sections 3.8.1.1.1 3.8.4.5.1, and 3.7.B.2.8, the allegation is judged to have merit and potential safety significance. Prompt remedial actions as delineated below should be implemented. No closing interview could be held regarding this allegation, because the allegation was received anonymously.
- 6. Actions Required: TUEC shall:
- 1. Perform an inspection of the as-built condition to confirm that adequate separation for all seismic Category I structures has been provided.
I
- 2. Provide the results of analyses which demonstrate that the presence l of rotofoam and other debris between all concrete structures (as ;
determined by inspections of the as-built conditions) does not result' I in any significant increase in seismic response or alter the dynamic response characteristics of the Category I structures, components, and piping when compared with the results of the original analyses. l l K-77 l
1 i 1. Allegation Category: Civil and Structural 12, Concrete Construction and Deficiencies / Tolerances
- 2. Allegation Number: AC-29
- i
- 3. Characterization: It is alleged that a spillway pillar, span, or column was erected 75 degrees to 80 degrees offset and that reinforcing steel 4
was omitted from a concrete wall.
- 4. Assessment of Safety Significance: There are two spillways at Comanche Peak Steam Electric Station (CPSES). One, the service water discharge spillway, is located near the safe shutdown impoundment (SSI); the other is located at the Squaw Creek Dam. The alleger stated that the construc-tion in question took place some time between 1976 and 1977. The spillway at Squaw Creek Dam was constructed between August 1976 and January 1977, j
so it was considered to be the spillway in question. The spillway at l Squaw Creek Dam, however, does not_have a span, column or pillar. There-fore, on August 3,1984, the NRC Technical Review Team (TRT) interviewed
- the alleger to clarify this allegation.
From the interview, the TRT learned that the spillway pillar, span, or
- column to which the alleger referred was located in the Service Outlet
' Structure below the Squaw Creek Dam Spillway, which does have a suspended structure and supports that could be described as a span and pillars.
The TRT inspected the service outlet structure at the Squaw Creek Dam Spillway and found no evidence of any spillway pillar, span, or column which was erected 75 to 80 degrees offset. The TRT also determined that the general configuration of the structure was consistent with that shown 4 on the following drawings: FN-SCR-37 FN-SCR-48 FN-SCR-39 FN-SCR-49 FN-SCR-40 FN-SCR-71 l FN-SCR-42 FN-SCR-72 FN-SCR-44 The TRT learned during an interview with the alleger.that the allegation concerning the 6-foot by 6-foot concrete wall ares of the Safeguards ' Building, which allegedly had no reinforcement placed around a pipe approximately 24-inches wide, was incorrect. (Refer to Civil and j Structural Category 6, Allegation. AC-30.) The alleger identified the
- 6-foot by 6-foot concrete wall area as located in a structure near the ,
i Squaw Creek Dam Spillway. 4 The TRT inspected.the structures located near the Squaw Creek Dam spillway and found two structures with a 2- to 3-foot-diameter pipe
- surrounded by reinforced concrete. One of these was the outlet works
,, conduit section; the other was the return pump station. The TRT examined i
141 concrete placement cards associated with these two structures. 1 4 K-79
The TRT determined that the conduit section was placed between June 27 and November 17, 1975, and the return pump station section was placed between March 31, 1976 and February 10, 1977. Because the alleger's j employment on the project began in 1976, the TRT concludedThere that the are two allegation, if valid, concerned the return pump station. 24-inch steel pipes in the return pump station which pass through a concrete wall. The TRT reviewed reinforcement drawings (FN-PS-35 and FN-PS-36) for the wall at the return pump station and found that the wall section surrounding the pipe was designed to have the following reinforcement:
- a. Eight No. 5 diagonal bars at the inside face
- b. Eight No. 7 diagonal bars at the outside face
- c. Ten No. 7 vertical bars at the outside face
- d. Ten No. 5 vertical bars at the inside face
- e. Ten No. 7 dowels (lap spliced with item c)
- f. Eight No. 5 dowels (lap spliced with item d)
The walls of the return pump station were placed on June 21, 1976. The TRT examined the pertinent concrete placement card. It contained the required two signatures certifying that the reinforcement was correctly placed prior to concrete placement.
- 5. Conclusions and Staff Position: Since the structures at which the alleged construction deficiencies occurred are categorized as nonsafety related (FSAR Volume IV Section 3.2), the allegation is judged by the TRT to have no safety significance. Furthermore, the TRT concludes that the first part of the allegation is not valid because a structure that was constructed at 75 degrees to 80 degrees offset from the intended geometry could not be accepted by inspection personnel without detection of such a significant deviation. Field inspection by the TRT indicates correct alignment.
The TRT further concludes that the second part of the allegation is not valid because the concrete placement card indicates that the reinforcement was placed. Additional evidence is provided by the fact that the portion of the wall surrounding the 24-inch pipes has been subjected to the maximum static load stress for which it was designed. The soil pressure has been in place and acting upon the wall for several years, and the reservoir was completely filled by water pumped through two 24-inch diameter pipes passing through the wall; therefore, this portion of the wall has also been subjected to whatever vibratory loads may be imparted to the wall by the pumping operation. Inspection by the TRT revealed no distress in the wall, and the structural integrity of the wall was observed to be intact. Accordingly, this allegation has neither safety significance ..or generic implications. The TRT informed the alleger by letter of its disposition of this allegation.
- 6. Actions Required: None.
K-80 i
- 1. Allegation Category: Civil and Structural 13, Cracks in Concrete Beneath the Reactor Vessel
- 2. Allegation Number: AC-44
- 3. Characterization: It is alleged that detrimental cracks exist in the concrete pad at the bottom of the reactor vessel.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) did not initially attempt to contact the alleger because the allegation was sufficiently clear to the TRT to proceed with the investigation.
The existence of these cracks is documented in Nonconformance Report (NCR) C-650. The cracks are in a lift of concrete near the bottom of the reactor vessel, but not in the basemat. The TRT examined concrete placement pack-age 101-2812-001 and NCR C-650, and found no documented violations of specifications during the concrete placement, which occurred on March 21, 1977. Subsequent to placement, however, vertical cracks occurred that extended horizontally to the edge of the reactor cavity. The Gibbs & Hill (G&H) design engineer stated on May 11, 1977, that he found the cracks during his investigation. He attributed the cracks to the mass, configura-tion, and formwork on the interior circumferential face, all of which precluded normal shrinkage, and stated that the cracks were of no struc-tural significance. An NRC Region IV structural engineer also presented his evaluation of the cracks during the ASLB hearing conducted on June 9, 1982. The TRT reviewed this testimony, along with that of G&H engineers given on June 7 and and 8, 1982, and agreed with the assessment contained in their testimony. The doughnut shape of the concrete section and the rigid form in the opening made it virtually impossible to avoid cracking if the entire section was placed in one pour, as it was. However, the structure was adequately reinforced so that the cracks would not impair structural behavior and capacity. The cracks have been repaired at the surface with epoxy resin for operational, rather than structural, reasons. The TRT inspected the concrete and found it to be in excellent condition. The TRT review of the design indicated that the concrete section was originally designed as two sections, with construction joints at the locations where the cracks occurred. The contractor was given the option of placing the concrete either in two sections with construction joints, or in one section without joints. The cracks that formed were not greatly different from the construction joint which would have been present if the two placement option had been adopted; thus, the concrete in place essentially conformed with the original design. One of the cracks was near the mid-span of a deep beam spanning a 20-foot cavity. Reinforced concrete beams must crack in the bottom tensile zone when load is applied. If flexural stresses were kept below the tensile strength of concrete, less than 20 percent of the strength of the steel would be utilized. In design, the reinforcement is distributed so that the cracks are numerous and very narrow, both for the sake of appearance and to prevent corrosion of the steel. The occurrence of a pre-existing K-81
crack merely changes the distribution of cracks; the total width of the cracks in the tensile zone remains unchanged. A crack in the upper com-pressive stress zone closes when load is applied and is rendered innocuous. The beam section must also be capable of carrying shear stresses. The cracks observed should not produce a critical situation because shear stresses are low near midspan and because crack planes are normally irregular so that aggregate interlock, particularly in the tightly closed compressive zone, resists shear stress. The biggest defense against shear, however, is the fact that the concrete was heavily over-reinforced. The critical load condition is not the static load condition, nor even the earthquake condition, but the differential pressure resulting from a postulated accident condition. For this condition all the load is carried by the steel, with no credit given.to the concrete, and the presence of cracks in the concrete is immaterial. The design of the The section was controlled by thickness requirements for shielding. section was thicker and, therefore, stronger than required to carry the loads. The cracks did not make the steel vulnerable to corrosion because the upper surface, which provides the most likely ingress for water, is sealed, and the bottom surface is in a dry environment.
- 5. Conclusion and Staff Positions: Although the a' legation is correct in citing the existence of cracks, it is not cor'ect in _ imputing detrimental structural consequences to them. The safety t le structure is not adversely affected by the cracks. Accordingly, this allegation has neither safety significance nor generic implications.
The TRT held a closing interview with the alleger, who was satisfied with the TRT's disposition of this allegation.
- 6. Actions Required: None.
K-82 1
- 1. Allegation Category: Civil and Structural 14, Control Room Area Deficiencies
- 2. Allegation Number: AE-17
- 3. Characterization: It is alleged that the field run conduit, the drywall, and the lighting installed in the area above the ceiling panels in the control room are classified as non-seismic and are supported only by wires and that these items may fall as a result of a seismic event.
- 4. Assessment of Allegation: The NRC Technical Review Team (TRT) did not initially attempt to contact the alleger because the allegation was sufficiently clear to allow the TRT to proceed with its investigation.
The TRT electrical group reviewed the electrical aspects of this allegation. (See Electrical and Instrumentation Category 4.) The Civil and Mechanical group of the TRT evaluated the seismic aspects of this allegation. General Design Criteria No. 19 requires that safe occupancy of the control room during abnormal conditions be provided for in its design. The Comanche Peak Steam Electric Station (CPSES) control room is in a seismic Category I structure, with certain seismic Category II and nonseismic components located in the ceiling. Seismic Category I refers to those systems or components which must remain functional in the event of an earthquake. Seismic Category II refers to those systems or components whose continued functioning is not required, but whose failure could reduce the functioning of any seismic Category I system or component (as defined in Raulatory Guide 1.29) to an unacceptable level or could result in an incapacitating injury to occupants of the control room. Seismic Category II systems or components are, therefore, designed and constructed so that a Safe Shutdown Earthquake (SSE) will not cause such failure or injury. In assessing this allegation, the TRT reviewed the CPSES nonsafety-related conduit, lighting fixtures, and the suspended ceilings installed in the control room. Three types of suspended ceiling exist in the control room: drywall, louvered, and acoustical. The following list designates those ceiling elements present in the control room and their seismic category designation:
- 1. Heating, Ventilating and Air Conditioning - Seismic Category I
- 2. Safety-related Conduits - Seismic Category I
- 3. Nonsafety-related Conduits - Seismic Category II
- 4. Lighting Fixtures - Seismic Category II
- 5. Sloping Suspended Drywall Ceiling - Nonseismic
- 6. Acoustical Suspended Ceiling - Nonseismic
- 7. Louvered Suspended Ceiling - Nonseismic The TRT also examined the control room ceiling system and pertinent design drawings and met with cognizant Texas Utilities Electric Company (TVEC) engineers on July 31, 1984, to discuss the specific seismic analyses per-formed for the ceiling elements. In addition, the TRT held a conference call on August 1, 1984, with principal Gibbs & Hill (G&H) design engineers K-83
(at whic'i TUEC representatives were present) to discuss the design and calculation procedures for the ceiling elements. The TRT determined that none of the suspended ceiling elements were con-sidered to be either seismic Category I or II; however, TUEC had modified the sloping suspended drywall to add more support. G&H could not provide backup calculations to support this modification, nor could TUEC provide justification for their position that the remaining suspended ceiling elements (i.e., the louvered and acoustic elements) would not fall and cause an incapacitating injury to operating personnel. This would indi-cate failure of the quality assurance program to ensure that applicable provisions of Regulatory Guide 1.29 were fully met. The TRT requested backup calculations for the sloping suspended drywall. TUEC provided the calculations on August 3, 1984, along with the calcula-tion packages for the lighting fixtures, the nonsafety-related conduits larger than 2 inches in diameter, and the safety-related conduit. The TRT reviewed these calculations, except those for the safety-related conduit, since they were designated as seismic Category I and therefore were excluded from the scope of this review. The TRT found that nonsafety-related conduits that were less than or equal to 2 inches in diameter were not supported by redundant seismic Category II cable restraints. The TRT also verified the adequacy of calculations for the nonsafety-related conduits larger than 2 inches in diameter. The TRT found that the G&H calculations were based on the equivalent static load method, which involves multiplication of the dead weight of an item by an appropriate seismic acceleration coefficient. This equivalent static load calculation did not take into account the influence from the adjoining suspended ceilings on the calcubted response. This was signif-icant because redundant cable supp ets were not provided for the suspended louvered and acoustical ceilings, and the impact from the accelerations of the lighting fixtures was not considered in any analysis. The ceiling, as a whole, manifested a more complex configuration than that assumed in the equivalent static load analysis in that the effects from adjoining sus-pended ceilings were not considered. A justification based on the seismic response characteristics of the entire ceiling, which would account for the frequency content and amplification characteristics of the seismic motions, as represented by floor response spectra, is required to justify the value of the seismic acceleration coefficient used.
- 5. Conclusions and Staff Position: The TRT found that not all items in the Control Room ceiling fall under the seismic Category I or II designation.
Specifically, these items are the suspended drywall, acoustical, and louvered ceilings. These components, designated as nonseismic, do not-satisfy the provisions of Regulatory Guide 1.29, since they were not designed to accommodate seismic effects. Nonsafety-re, lated conduits that are 2 inches in diameter and less also were not designed to accommodate seismic effects. TUEC presented no evidence which showed that the effect of failure of these items had been considered. K-84
1 The TRT concludes that calculations supporting the seismic Category II lighting fixtures do not adequately reflect the rotational interaction with the nonseismic items. In addition, the fundamental frequencies of the supported masses were not calculated to determine the influence of the seismic response spectrum at the control room ceiling elevation. The individual who made the allegation discussed above will be contacted by the TRT upon resolution of this issue to inform him of the action taken.
- 6. Actions Required: TUEC shall provide:
- 1. The results of seismic analysis which demonstrate that the nonseismic items in the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. An evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems.
- 3. Verification that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
- 4. The results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
- 5. The results of an analysis which demonstrate that the foregoing problems are not applicable to other Category II and nonseismic structures, systems, and components elsewhere in the plant.
K-85
l l l l
- 1. Allegation Category: Civil and Structural 15, Rebar Improperly Drilled
- 2. Allegation Number: AC-13, AC-14, AC-15, AC-18 and AC-40
- 3. Characterization: It is alleged that undocumented and unauthorized holes were drilled through reinforcing steel (rebar). The issue includes allegations relating to:
- a. the loan of rebar drills without proper documentation (AC-13),
- b. the unauthorized cutting of rebar in non-specific locations (AC-14, AC-18, AC-40), and
- c. the unauthorized cutting of rebar used in the installation of the trolley process aisle rails in the Fuel Handling Building (AC-15).
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) contacted the individual who made allegations AC-13 and AC-14 to clarify his concerns. The TRT did not initially attempt to contact the individuals who made allegations AC-18, AC-40, and AC-15.
- a. AC-13 concerns the loan of rebar drills allegedly used for the unauthorized cutting of rebar. During the NRC investigation of this matter, the NRC Office of Investigation (OI) interviewed nine individuals alleged to have knowledge of unauthorized cutting of rebar. These individuals provided sworn statements denying any knowledge of this activity. These statements are a part of OI Report A4-83-005 (May 20, 1983), which concludes that "there was no testimony received indicating that holes were drilled or rebar was cut without proper documentation, and no evidence was found to contradict the testimony of these individuals." One instance of possible unauthorized cutting of rebar is discussed in a supplement to the OI report (September 7, 1983). This instance is discussed below in relation to allegation AC-15.
Because the alleger did not specifically identify who made unauthorized cuts of rebar, or where this cutting took place, the TRT attempted to quantify the amount of rebar that allegedly was cut without authorization. In discussions with the TRT, the alleger estimated that approximately five percent of the diamond core drill bits ordered by him were used in an unauthorized manner. He further estimated that one drill could be used to cut up to five rebars, depending upon the extent of cutting required. Although he could not be specific as to how many drills he ordered, the alleger thought that the number would be in the thousands. The NRC Region IV Investigation of this issue indicated that 415 diamond core drill bits were purchased during the period in question (IE Report 83-27). Using the actual number of drill bits purchased, together with the 1 information provided by the alleger, the TRT estimated that there ' could be approximately 100 alleged unauthorized rebar cuts. Considering the large amount of reinforcing steel used in the plant, and the fact that the structures consist primarily of heavily reinforced concrete walls and slabs, the TRT determined that, if such K-87 l 1
l unauthorized rebar cutting occurred, the amount involved would have an inconsequential effect on the safety of the structures. j i
- b. Allegations AC-14, AC-18 and AC-40 also raise nuestions regarding the unauthorized cutting of rebar, but do not idantify specific locations. During the course of the NRC Region IV investigation of this matter, the alleger provided a log bo9k which, it was reported, would identify the unauthorized and undocumented rebar cutting.
However, the Region IV inspector could not identify one rebar cut listed in the log that was not authorized. The TRT also reviewed the log and came to the same conclusion. In discussing this matter with the TRT, the alleger confirmed that there was documentation supporting "ninety nine and three quarter percent" of the rebar cuts identified in the log. As part of Report 83-27, the NRC Region IV investigator traced 32 authorizations, approximately half of the documents noted in the log for the rebar cutting. He found that in all cases rebar cuts were properly identified on a design change authorization (DCA) or on a component modification card (CMC). In addition, the rebar cuts were traced to and identified on specific building structural drawings, with the corresponding authorizing document number. The TRT reviewed 10 CMCs ] and confirmed the findings of the Region IV investigation. In reviewing authorizations in the log, the TRT noted that certain CMCs involved a number of rebar cuts in one area, and selected these for review. In one case, 7 different CMCs (3307, 3664, 3665, 3666, 3667, 3668 and 3669) seemed to pertain to one area and accounted for 68 rebar cuts. Upon reviewing the documentation, the TRT found that these cuts were made in a tunnel area in the Fuel Building. (The alleger identified this as a location where a large number of rebar was cut.) However, the 68 cuts were arranged such that only 9 bars actually had been cut. In another case, the log in'dicated 25 rebar cuts pertaining to CMC 00979. In this case, the TRT determined that all the cuts were made on one reinforcing bar in a support beam. Finally, the log indicated eight rebar cuts pertaining to CMC 3022. Once again, these eight cuts were to one bar in a support beam. All cuts were made in accordance with the rebar cutting criteria provided by Gibbs & Hill. These examples also illustrate the point that a large number of rebar cuts recorded are not necessarily synonymous with an identical number of rebar actually being cut. In all cases, one bar was cut a number of times, but adjacent bars were not. Thus, the cuts were arranged to minimize the overall effect on the strength of the structure. The TRT estimates that approximately 335 rebar cuts are indicated in the alleger's log. Discussions with the alleger revealed that he believes he cut approximately five percent more rebar than was authorized, a number that corresponds to approximately 17 unauthorized rebar cuts. As noted earlier, such a number would have little effect on the safety of the structures. K-88
, - , - - , - ,- - ~ , , , ,
l l I
- c. Allegation AC-15 identifies a specific instance of the possible unauthorized cutting of rebar. In this case, a'former Brown & Root employee stated he possibly drilled holes through rebar in a concrete floor without a component modification card (CMC) or a design change ,
authorization (DCA). He explained that in January 1983 he drilled approximately 10 holes about 9 inches deep while installing 22 metal plates with a core drill. He said the metal plates were used to secure the trolley process aisle rails located on the 810-foot, 6-inch floor level in Room 252 of the Fuel Handling Building. The TRT inspected the trolley process aisle rails and its anchoring system and observed no violations of project drawings or specifications. The TRT reviewed the reinforcement drawings (2323-S-0800 and 2323-5-0820) for the Fuel Handling Building to determine the location of rebar. The drawing showed three layers of reinforcement in the upper part of the mat, which consisted of a No. 18 bar running in the east-west direction, in the first and third layers, and a No. 11 bar running in the north-south direction, in the second layer. The review of the reinforcement drawings (2323-5-0800 and 2323-5-0820) revealed that the layout of the east-west reinforcement and the j trolley process aisle rails was such that only one bar of the east-west reinforcement could be cut by drilling holes for rail anchors. However, if 9-inch holes were drilled, both layers of the No. 18 reinforcing bar would be cut. De' sign Change Authorization (DCA) No. 7401 was written for authorization to cut the uppermost No. 18 bar at only one rail, but it did not reference the authorization to cut the lowermost No. 18 bar. The DCA (No. 7041) also stated that the expan-sion bolts and baseplates could be moved in the east-west direction to avoid interference with the No. 11 reinforcement running in the north-south direction. The information described in DCA No. 7041 was substantiated by Gibbs & Hill calculations. The DCA approval was based on the understanding that only the uppermost No. 18 reinforce-ment would be cut. If the 10 holes were actually drilled 9 inches deep, then the allegation that reinforcement was cut without proper authorization may be valid. The DCA indicated that the holes were drilled to accommodate 1/2-inch Hilti bolts, which require a minimum embedment of 5-1/2 inches (as noted in Fig. 39, Sh. 5 of 5, attached to DCA-7041). Since there was no need to drill the holes deeper than 5-1/2 inches, the alleger may not be correct in stating that the holes were drilled 9 inches deep. Although the allegations discussed above, with the exception of AC-15 which requires further action, cannot be substantiated, the fact that such allegations were made indicated that there was no effective quality assurance program to oversee the issuance and use of diamond core drill bits. The TRT interviewed the individual concerned about the loan of rebar drills without proper documentation and the unauthorized cutting of rebar at nonspecific locations to inform him of the TRT's finding. This individual did not agree with certain TRT findings. In K-89 l
- \
1 J i I 1 particular, the alleger felt that the TRT's estimate of approximately l 120 unauthorized rebar cuts was much too low. He believes that the i number of drill bits ordered by him was in the thousands and that as much as 20 percent of the drill bits may have been used in an !
! unauthorized manner. It was also his opinion that the unauthorized cutting of rebar was not limited to his period of employment, but occurred for the duration of the project.
i As a result of these additional discussions with the alleger, the TRT searched TUEC's files relating to the purchase of diamond drill bits li and found that 1170 drill bits were purchased between January 13, 1978 and January 14, 1980. This number is more in agreement with the alleger's assessment and is higher than the previously reported number of 415 (IE Report 83-27). The TRT also found that there were a total of 3368 drill bits ordered from one manufacturer between , January 13, 1978 and March 18, 1983. After this period, other j manufacturers supplied the drill bits. Based on the usage through March 10, 1983, the TRT estimates that approximately 5000 diamond
! drill bits have been used to date on the project. Assuming that 20 ; percent of these drill bits were used in an unauthorized manner and ; that each drill bit could be used to cut up to five rebars, the TRT 4
estimates that there could be approximately 5000 alleged unauthorized ; rebar cuts. ! The TRT estimated that, depending upon the average length of_rebar assumed, there are approximately 800,000 to 1,200,000 bars installed in all of the concrete structures. Thus, if 5000 bars were cut without authorization, they would represent approximately 0.6% of the-
; total rebar in the plant. Even if all 5000 drill bits were used in an unauthorized manner it still would only represent 3% of the total rebar in the plant. Thus the percentage of rebar that could have been cut witnout proper authorization is low. Since no information was supplied to the contrary, the TRT assumed that these unauthorized J cuts, if they did occur, were scattered throughout the plant and not j concentrated in one localized area. In addition, as noted earlier, a I large number of rebar cuts are not necessarily synonymous with an I identical number of rebar actually being cut.' It is also noted that nuclear structures are very conservatively designed. In addition to the conservative loads,-load combinations,.and safety factors utilized-in the design, it is the common practice of the design engineer to specify 5 to 10 percent more rebar than is actually required by his calculations. This occurs because it is difficult to i
} obtain the exact area of reinforcement required using standard bar sizes and standard bar spacing. The area of reinforcement is selected from charts which show-the area provided for each bar size
' at a given spacing. Rather than underdesigning, the designer selects i
an area of reinforcement from the charts which is higher than that which is actually required. In addition,-because critical structures
~
I i contain a large number of bars, they are not generally vulnerable to the random cutting of a small number of bars. ! 5. Conclusion and Staff Positions: The TRT concludes that allegations AC-13,
' AC-14, AC-18 and AC-40 have no structural safety significance.
K-90 i l
i
- a. The allegations were not specific as to who made unauthorized cuts of rebar or where the cuts took place.
- b. The number of unauthorized rebar cuts alleged, if true, would have an inconsequential effect on the safety of the structures.
However, the results of these evaluations will be further assessed as a part of the programmatic review concerning procedures addressed under QA/QC Category 6, "QC Inspection." Therefore, the final acceptability of these evaluations will be predicated on the satisfactory results of the programmatic review of this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER. Allegation AC-15 will remain open until the information requested of Texas Utilities Electric Company (TUEC) in " Actions Required" is provided. The TRT attempted to contact the individuals who made allegations AC-18 and AC-15 to inform them of the TRT's findings. The individual who made allegation AC-18 will be informed of the TRT's findings by letter. One of the two individuals involved in allegation AC-15 cannot be located; the TRT is still attempting to contact the other. The TRT also contacted the individual who made allegations AC-13, AC-14, and AC-40 to discuss the TRT's findings pertaining to the concerns he raised in the first closure interview. An interview.was arranged; however, later the alleger indicated he did not want to meet with the TRT. A letter will be sent to him informing him of the TRT's findingr.
- 6. Actions Required: TUEC shall provide:
- 1. Information to demonstrate that only the No.18 reinforcing steel in the first layer was cut, or
- 2. Design calculations to demonstrate that structural integrity is maintained if the No. 18 reinforcing steel on both the first and third layers was cut.
l K-91
- 1. Allegation Category: Civil and Structural 16 Excavation and Backfill
- 2. Allegation Number: AQ-64
- 3. Characterization: It is alleged that overexcavation and improper fill under the Unit 1 Containment Building could invalidate the expected seis-mic response of the foundation due to the change in properties resulting from the removal of in-situ material.
t
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) did not initially attempt to contact the alleger because the allegation was sufficiently clear to allow the TRT to proceed with its investigation.
During an investigation conducted in 1984, the NRC Office of Investigation (01) interviewed the alleger (84-006, 3/7/84, A-7) and reference was made to overexcavation and improper repairs in the foundation rock for the Unit 1 Containment Building. The alleger stated that the excavation was erroneously made 6 to 8 feet too deep and that upon realization of the i error, the repair technique was simply to throw the loose rock back into the excavation and fill it in with concrete. The TRT reviewed NRC inspection reports, the FSAR, and the Atomic Safety and Licensing Board (ASLB) hearing transcript, where this concern was the subject of contention No. 7 and was admitted into the hearing on June 16, 1980. By order of March 5, 1982, the ASLB granted summary disposition of con-tention No. 7, based on the finding that no genuine issue as to any material fact was shown by any of the filings. The TRT also reviewed the affidavits and statements filed by TUEC and by the NRC in support of the motion for summary disposition. These documents adequately describe rock overbreak, accompanying fissures, and subsequent repairs. Affected areas were backfilled with concrete having a minimum compressive strength of 2,500 pounds per square inch at 28 days, or were grouted to maintain continuity of the competent rock in which fissures were identified. The TRT reviewed the procedures utilized to replace fractured rock with dental concrete and. to grout surrounding fissures and the accompanying compres-sive test results. The TRT found that FSAR figures 2.5.4-33a through 2.5.4-35 are maps of the excavation showing the location of fractures and the extent of dental concrete backfill. These figures showed that the area of overexcavation represented a small portion of the entire excavated area. FSAR figure 2.5.4-37, sheets ?. through 21, showed photographs of the excavated walls. The TRT interviewed the NRC inspector who was present i during the excavation process and verified the conditions presented in the FSAR. The TRT independently evaluated the potential impact on the seismic re-sponse of the Unit 1 containment foundation due to the replacement of a limited amount of original rock with dental concrete from the standpcint of possible changes in foundation stiffness. Because of the facts that I (a) the dental concrete's behavior, stiffness, and structural strength were essentially identical to those of the natural rock replaced at the l K-93 l
site as indicated by the foundation report and (b) the area affected by the replacement work was relatively small (refer to FSAR Figures 2.5.4-33a through 2.5.4-35), the TRT determined that no appreciable impact on either the static or dynamic response characteristics of the foundation resulted from the overexcavation. An evaluation prepared by a geotechnical engi-neer in the NRC's Office of Nuclear Reactor Regulation supports this con-clusion. He evaluated the effects on static and dynamic foundation sta-bility of replacing undisturbed limestone and claystone foundation rock with dental concrete and concluded that the ability of the repaired foun-dation materials to withstand seismic disturbances had not been impaired.
- 5. Conclusion and Staff Positions: The TRT concludes that the overexcavation of a small portion of the Unit 1 Containment Building foundation and the subsequent replacement of the affected area with 2500 psi strength dental concrete and grout did not affect either the static or dynamic character-istics of the foundation. Therefore, the expected seismic response has not been invalidated as alleged. The excavation and repairs have had no safety impact upon foundation integrity. Accordingly, this allegation has neither safety significance nor generic implications.
The TRT has contacted the alleger to arrange an interview to inform him of the TRT's finding.
- 6. Actions Required: None.
l 1 K-94
- 1. Allegation Category: Civil and Structural 17, Concrete Sampling
- 2. Allegation Number: AQC-45
- 3. Characterization: It is alleged that personnel produced incorrect readings on concrete batch plant scales by leaning on the wires connecting the weighing hoppers to the scales.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) did not initially attempt to contact the alleger because the allegation was sufficiently clear to allow the TRT to proceed with its investigation.
The allegation refers to "some type of sampling machine that tells whether there are good samples or bad samples in the concrete." The information is attributed to a friend of the alleger's who was an equipment operator at the concrete batch plant. NRC Inspection Reports 50-445/83-01 and 50/446/83-03 clarified the allegation, indicating that the equipment oper-ator is reported to have " charged that some personnel biased the operation of the concrete batch plant scales by leaning on the wires connecting the scales to the sensors." i During the major construction phase of the project, the concrete plant consisted of two identical batching systems, one feeding a mixer drum and one batching directly into ready-mix trucks. Although the system with the self-contained mixer was removed prior to the TRT review, the other system was still in use at the time of the TRT review. The batching system in-cluded three mechanical lever dial scales, each controlled by a wire con-nected to the weighing hopper with which it is associated. Each wire entered the scale room over a pulley at the top of the room and ran down-ward vertically about 6 feet to the loading lever of the scale. The scale dials faced the control room, from which they were visible through a large window. An electronic sensor connected to the scale provided a digital readout of each scale reading in the control room. It was possible to decrease the scale reading by entering the scale room and horizontally deflecting the vertical wires. The TRT interviewed the concrete plant maintenance man who was present during most of the construction and who, at the time of the review, was also serving part time as operator because of the small amount of concrete production. He stated that the scale room was enclosed, well-illuminated, and provided with a large window so that all parts of the enclosure were visible from the control room, making it easier to prevent surreptitious tampering with the scale mechanisms. Thus, it would be obvious to everyone in the control room.and to many outside the control room if anyone opened the scale room door, entered, and deflected the wires. Tbc mintenance man was not aware that such an incident had ever occurred. Since it was not possible to rule out such tampering completely, the TRT investigated the potential consequences of such tampering. During a con-crete placement, a member of the TRT entered the scale room when all hoppers were loaded and deflected each scale wire as far as could be , conveniently done by one person. The scale readings were affected as follows K-95
Normal Deflected Reading Reading Aggregate scale 14,500 lbs 14,400 lbs Cement scale 3,150 3,050 Water scale 1,040 990 The aanner in which the scales are constructed makes possible only a de-crease in readings, not an increase, if the scales were tampered with as alleged. This arrangement rules out the most common allegation of fraud in concrete batch plants: inflation of the cement batch weight. If the cement scales were tampertd with, it would be necessary to add extra ce-ment to satisfy the stipulated batch weight. Moreover, the possible change in aggregate weight is within permitted tolerances and may be ignored. The only ingredient of concern is water. Tampering with the water scale could cause an extra five percent to be batched, which would increase the. water-cement ratio from 0.50 to 0.525. However, water is the one ingredi-ent in concrete whose abuse in batching is most easily detected in fresh concrete. A five percent increase of water would increase the slump by nearly 2 inches. The good control of slump, as verified by test data in the many concrete placement packages reviewed by the TRT, strongly indi-cated that there was no tampering with the water scale, the only scale vulnerable to the type of tampering alleged which would adversely affect safety. Thus, the evidence suggests either that tampering did not occur or that it was confined to scales where tampering would have either no effect or a beneficial effect on the concrete.
- 5. Conclusion and Staff Positions: The TRT interviewed relevant personnel, observed the layout of the scale room, conducted a demonstration of the tampering alleged, reviewed test data on freshly placed concrete, and examined two NRC inspection reports in evaluating this allegation. Based on its findings, the TRT concludes that the allegation can be neither veri-fied nor refuted. However, if tampering did occur, it was confined to scales where tampering would have either no effect or a beneficial effect on the concrete. Accordingly, this allegation has no structural safety significance. However, the results of this evaluation pertaining to QC controls at the batch plant will be further assessed as part of the over-all programmatic review concerning procedures addressed under QA/QC Cate-gory 6, "QC Inspection." Therefore, the final acceptability of this evalu-ation will be predicated on the satisfactory results of the programmatic review on this subject. Any adjustments to the existing conclusion of this evaluation resulting from the programmatic review will be reported in a supplement to this SSER.
The individual who made this allegation will be informed of the TRT's findings by letter.
- 6. Actions Required: None.
l
)
K-96
- 1. Allegation Category: Miscellaneous 1, Nuclear Fuel
- 2. Allegation Number- AM-?
- 3. Characterization: It is alleged that nuclear fuel was received prior to issuance of a special nuclear material (SNM) license.
- 4. Assessment of Safety Significance: The NRC Region IV (RIV) staff received this allegation by telephone in January 1983, from a member of the public.
The caller had friends working at Comanche Peak Steam Electric Station (CPSES) who claimed that fuel was received onsite before the NRC issued an SNM license. Despite requests from RIV for either more specific infurma-tion about the allegation or the identity of those making the allegation, the NP.C staff received no substantive information because the caller stated that the allegers feared they would lose their jobs. The NRC Technical Review Team (TRT) reviewed the SNM license for Texas Utilities Electric Company (TUEC), issued February 14, 1983, the " dummy fuel assembly" receipt package, and the first-fuel receipt package. During this review, the TRT found that TUEC received a " dummy fuel assembly" onsite on Dccember 15, 1982, at 9:30 a.m., as noted on Form RF0-201-1, " Fuel Receiving Record - Shipment Report." Form RF0-201-1 also shows that the first-fuel shipment was received onsite on May 4, 1983, at 1:45 a.m. The TRT interviewed both the Radiation Protection Engineer (RPE) and the Radiation Protection Supervisor (KPS) who stated that receipt of the
" dummy fuel assembly" was used as a training exercise to prepare for incoming shipments of fuel expected to arrive at CPSES. The RPS stated that when CPSES received the " dummy fuel assembly," fuel-receipt procedures were followed as if it were a "real case" shipment. Thus, it is conceiv-able that the alleger mistakenly assumed that the " dummy fuel assembly,"
which was received onsite prior to issuance of the SNM license, was actu-ally nuclear fuel.
- 5. Conclusion and Staff Positions: Based on a review of documents and forms, on interviews with the RPE and RPS, and on a review of an NRC Region IV interoffice memorandum assessing TUEC's program for receiving fuel, the TRT concludes that TUEC has an adequate program to receive fuel and did not actually receive fuel onsite prior to issuance of an SNM license on February 14, 1983. Accordingly, this allegation has neither safety significance nor generic implications.
The TRT was unable to learn the identity of the alleger during the inspec-tion; therefore, no followup interview with this alleger was possible.
- 6. Actions Required: None.
K-97
- 1. Allegation Category: Miscellaneous 2, Reactor Pressure Vessel
- 2. Allegation Numbers: AM-3 and AM-23b.
- 3. Characterization: Two allegations concerning the reactor pressure vessel (RPV) are characterized as follows:
- a. It is alleged that during hot functional testing (HFT) expansion caused the reactor pressure vessel reflective insulation (RPVRI) to come in contact with the concrete biological shield wall (AM-3).
- b. It is also alleged that the Unit 1 RPV is located 3/16 inch to the west of the north-south centerline through the containment building (AM-23b).
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) reviewed the Final Safety Analysis Report (FSAR) for verification of RPV reflective insulation quality class; background material in the NRC Inspection Report 50-445/83-34; 50-446/83-18 and the affidavit of Doyle Hunnicutt clarifying this report; Brown & Root (B&R) construction opera-tion traveler 35-1195; and Westinghouse (W) Field Change Notices (FCNs)
TBXM-10609, 10611, and 10612. The TRT also reviewed photographs of con-struction debris obstructions between the RPVRI and the biological shield and discussed all of these documents with the Texas Utilities Electric Company (TUEC) Mechanical Engineering Supervisor, the W project manager, and the TUEC heating, ventilation, and' air conditioning system (HVAC) startup group leader. The TRT also reviewed documentation for hot func-tional test (HFT) No. ICP-PT-5502 for any i'nformation concerning the allegations, but found no specific notations about them. The Comanche Peak Steam Electric Station (CPSES) FSAR, Volume XIV, Appendix 17-A, classifies the RPVRI as nonsafety-related. The support ring (a support channel) for the RPVRI is part of the RPVRI; therefore, it too is considered to be nonsafety related. The only function of this sup-port channel is to support the refective insulation. It is not needed for the safe shutdown of the plant but simply insulates the vessel. In this case it assumcd iraportance only because of its effect on adjacent safety-related structures or components. A Westinghouse FCN (TBXM-10609) dated September 27, 1983, documented an unacceptable condition which was identi-fied during hot functional testing. Actual air temperatures during HFT were 288*F near the reactor vessel flange versus 150 F maximum in the cool-ing annulus between the RPVRI and the shield wall. Similarly unacceptable temperatures were noted (150 F actual vs 135*F allowable) in the ex-core detector wells. TUEC letter TX5054 reportcd extreme temperatures up to 314 F. Further examination by W personnel revealed that cooling air flow was restricted by the RPVRI support channel because of constructicn debris between the RPVRI support channel rnd the steel-lined concrete biological shield wall and because of restriction by the support channel. The debris was removed by a remote technique and the gap was fiberoptically inspected. Instead of the nominal design gap (cold) distance of 7/8 inch between the support channel and the shield wall, this inspection identified cold air gaps as follows: a 7/8-inch air gap extending one quarter around (90 ) K-99
d r the annulus circumference, a 1/2 inch air gap extending one half (180*)' < around the circumference, and a 1/4-inch air gap extending around the remaining quarter (90') of the circumference. The TRT discussed the field change notice with the TUEC Project Mechanical Engineer and the Westinghouse Site Manager. Specifically, the staff asked { the reason fer there being too little space between the reactor vessel insulation support channel (which supports the reflective insulation) and 1 the biologict.1 shield wall. The TRT learned that the original Westinghouse specification required the support channel to be inside the insulation, i but TRANSCO Inc., the vendor, requested a design change to permit the sup-port channel to be placed outside of the insulation. Gibbs & Hill _Inc., the Architect Engineer, did not incorporate this change into the design nor did they consider the impact on the cooling of the reactor cavity; thus, there was_too little clearance between the outer circumference of the sup-port channel and the shield wall, which.resulted in restricted air flow and overh_ eating. l-During the HFT, TUEC identified and recorded the inadequate cooling as a
- test deficiency. On July 29, 1983, they reported the HFT deficiency (orally) to the NRC and formally reported the deficiency as a 10 CFR Part
- 50.55(e) item on August-25, 1983, completing their reporting in TUEC i
letter TXX-4054, dated September 26, 1983. To correct this deficiency TUEC. modified the insulation support channel by cutting holes in the top
- and bottom flanges of the channel to allow sufficient air-flow and heat j removal and to ensure proper operation of the ex-core detectors and pro-tection of the biological shield wall. TUEC made this modification in i accordance with W procedure MP 2.7.1/TBX dated October 1,1983. In j addition, the existing insulation seals and heat removal capacity were j improved. Further discussions with the TUEC engineer revealed that air j flow tests had been performed since the Unit 1 support channel was
' modified; however,. final results of these tests will not be known until the HFT is redone. The retest was-scheduled, completed, and is being-evaluated. The TRT found that the 50.55(e) report of the corrective action taken ! regarding this deficiency did not include determination of the underlying i cause of the deficiency. In addition, the report included no discussion of the effect on Unit 2 or how such a-deficiency could be prevented in ! Unit:2. However,'TUEC did fiberoptically-inspect Unit 2 for debris or a
~
similar gap, and found no problems. The TRT-determined that the areas- _ where debris entered the. gap have been sealed in'both Units 1 and 2,.and TUEC ~ anticipates no further problems with debris. The-TRT also evaluated the allegation concerning improper placement of the Unit 1 RPV by reviewing RPV construction operation traveler (C0T) No. 35-1195 and the field survey note, " Final Setting Alignment,"" dated June 29, 1978. In addition,'the TRT reviewed the W " Mechanical Service Manual," l
; whichincludesprocedures,forsettingtheRPV7withtheWprojectmanager to determine the importance of the location of the RPV. relative to the centerline of the _ Containment Building. The W manual states that align- _l l ment with other nuclear steam supply system'(NSSS) components is the most i l
' K-100 d l l
critical location factor, and alignment with the center of the Containment Building is of secondary importance. The Unit 1 RPV is alleged to be misaligned by 3/16-inch; however, opera-tion No. 7 of COT No. 35-1195 shows a setting tolerance of i 1/4 inch. The alleged 3/16-inch misalignment is within this tolerance. Attachment 35-1195-MCP-1 of COT No. 35-1195, " Alignment Location Record - Alignment Final Set," shows the maximum RPV deviation from the north-south and east-west centerlines of the Unit 1 Containment Building to be 0.003 inches which is within the specified tolerance. The TRT reviewed installation records and determined that the critical relationship of the NSSS compo-nents to each other, as well as to the Containment Building centerlines (N-S, E-W), was accurately maintained during installation of the reactor pressure vessel.
- 5. Conclusion and Staff Positions: Based cn review of documentation and discussions, the TRT concludes that the RPVRI did make contact with con-struction debris, but did not contact the steel-lined concrete biological shield wall as specifically alleged. During fiberoptic inspection, TUEC personnel observed no visible damage to the reflective insulation, and all corrective modifications were accomplished and accepted in accordance with procedure MP 2.7.1-TBX-3 and FCNs TBXM-10609, 10611 and 10612.
The allegation, as specifically stated, cannot be substantiated, although it does have some merit because an unsatisfactory condition did exist in that the reflective insulation made contact with debris. However, this allegation has both safety significance and generic implications because of peripheral issues; i.e. , failure to assure that proper design changes were communicated between organizations, failure to determine and report the un-derlying cause of a significant deficiency, and failure to ensure a proper gap between the support channel and shield wall when the vessel was set. The TRT also concludes that the RPV is set within the design location tolerance. Therefore, this allegation is not substantiated and has neither safety significance nor generic implications. The TRT is unable to interview the alleger to provide its findings and conclusions because the identity of the alleger is unknown. The TRT could not identify the alleger responsible for allegation AM-3 because an anonymous alleger called the Dallas Times-Herald. The identity of the alleger of AM-23 is unknown because the person who received the allegation did not record the alleger's name and no longer remembers it.
- 6. Actions Required: TUEC shall:
- 1. Review their procedures for approval of design changes to nonnuclear safety-related equipment, such as the RPVRI, ard make revisions as necessary to ensure that such design changes do not adversely affect safety-related systems.
K-101
- 2. Review procedures for reporting significant design / construction deficiencies, pursuant to 10 CFR Part 50.55(e), and make changes as necessary to ensure that complete evaluations are specified.
- 3. Provide analysis which verifies that the cooling flow in the annulus between the RPVRI and the shield wall of Unit 2 is adequate for the as-built condition.
- 4. Verify during Unit 1 hot functional testing that completed modifica-tions to the RPVRI support ring now allow adequate cooling air flow.*
The TRT notes that control of debris in critical spacings between compo-nents and/or structures was identified as an issue both in the investiga-tion of this allegation and in the civil and structural area (Item II.c,
" Maintenance of Air Gap Between Concrete Structure"), contained in Darrell G. Eisenhut's September 18, 1984, letter to TUEC (Attachment 3). Accord-ingly, TUEC shall also:
- 1. Identify areas in the plant with spacing between components and/or structures that are necessary for proper functioning of safety-related components, systems, or structures in which unwanted debris may collect and be undetected or be difficult to remove.
- 2. Inspect the areas and spaces identified and remove debris.
- 3. Institute a program to minimize the collection of debris in critical spaces and periodically reinspect these spaces and remove any debris which may be present.
i
*The test has been completed. However, TUEC's analysis of test results is still underway.
K-102
-__-._--..__._._----.___.-____r_
- 1. Allegation Category: Miscellaneous 3, PSAR Errors
- 2. Allegation Number: AM-4
- 3. Characterization: It is alleged that Sections 10.2-11 and 10.2-12 of Volume VII of the Comanche Peak Preliminary Safety Analysis Report (PSAR) contain errors.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) could-not contact the alleger because the alleger's telephone number and address were unknown and relatives would not-provide the TRT with this information.
The TRT determined that this allegation refers specifically to PSAR Sec- ' tion 10.2-11 and 10.2-12; however, in the alleger's statement to the NRC Region IV Office of Investigation, the terminology "FSAR" (Final Safety Analysis Report) was frequently used when the term "PSAR" apparently was intended. The alleged errors refer to turbine missile energy calculations in PSAR Volume VII, Sections 10.2-11 and 10.2-12. Sections 10.2-11 and 10.2-12 of the FSAR do not pertain to turbine missile energy. The referenced PSAR sections contain a bounding design analysis calcu-lationusedtodeterminethemaximumenergyavajlable-ifagurbine failure occurs. Using the equation, E = = w/g v the calculated value should be 18.27 x 10 6 ft-lbs;hoNver,mvthe PSAR erfoneo,usly stated the calculated value as 18.27 x 10 ft-~1bs. It appears that the exponent was dropped because of a typographical error. The PSAR reflects the preliminary plant design, but it is the FSAR that reflects the final plant design and safety analysis. Sections 10.2, 3.5.1.3, and 1.3 (Table 1,3-2, page 27) of the FSAR contain the final design and analysis for the turbine generator. These sections do not contain the alleged erroneous calculation. The FSAR refers to "Allis-Chalmers P.S. Inc., ER-503, ' Turbine Missile Analysis for 1800 t/ min NSTG with 44-inch Last Stage Blades,' July 1985." The analysis in this docu-ment is specifically applicable to the turbine generator which was actually purchased. The TRT reviewed the Allis-Chalmers analysis in the FSAR and determined that it corrected the turbine missile energy calculations in the PSAR. The TRT found no errors in the Allis-Chalmers analysis of turbine. missile probability, and concurred with the Texas Utilities Electric Company - (TUEC) conclusion that even if missiles were generated, they would be-contained by the low pressure turbine casing. Although the alleged error appeared to be a typographical error, the TRT randomly selected calculations from other FSAR sections and paragraphs (3.5A-1, 3.5-16A, and 10.3-5) for review and found no additional errors.
- 5. Conclusion and Staff Positions: The TRT reviewed the PSAR and determined that there was an error in the PSAR as alleged, but that it appeared to be 1 a typographical error. However, a review of the FSAR showed that the i alleged error in the preliminary design calculation of the PSAR had been- I corrected in the FSAR. The NRC staff also randomly reviewed other calcu- I lations in the FSAR and found that they were free of error. Additionally, ]
K-103 q pg -=, w- y y m = w e ---.yv e <
I the FSAR has been reviewed in detail by the NRC Office of Nuclear Reactor Ragulation (NRR) as part of the licensingAccordingly, process and this this included review of the turbine missile analysis. allegation has neither safety significance nor generic implications. The TRT contacted the alleger to discuss its findings and conclusions; however, the alleger declined to participate in the planned followup interview. A letter was sent to the alleger on January 22, 1985, re-y questing that he reconsider participating in an interview with the TRT; however, there has been no response to this letter.
! 6. Actions Required: hone.
l 1 I 1 K-104 ] 1
b 4
- 1. Allegation Category: Miscellaneous 4, Radioactive Material Release
+
- 2. Allegation Number: AM-5
- 3. Characterization: It is alleged that someone " threw'something radioactive i
in the lake," that is, in the Comanche Peak Reservoir, sometime between j September and November 1978, and that this material may have been tritium. l 4
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) reviewed the background material contained in NRC Region IV Inspection l Report No. 50-445/79-22. 50-446/79'21 which documents an inspection con-ducted September 7 and 12 and October 11, 16, 18 and 20, 1979, which followed up on allegations that appeared in the University of Texas at Arlington (UTAl newspaper, the " Shorthorn," on Wednesday, July 18, 1979.
j This inspection concluded that "this allegation appears to have no merit." , The TRT also reviewed an NRC Office of Investigation transcript of inter- t view No. 84-006, dated March 7, 1984. From this document, the TRT learned that the alleger stated that he was looking through his turbine book and " he saw that they used tritium gas to inspect for leaks. The alleger con-nected this with a story that another worker had told him, i.e., that one night the other worker was working down there when they threw something radioactive in the lake. The alleger in turn contacted the NRC to provide this information.
! The TRT interviewed both the TUEC Radiation Protection Engineer (RPE) and i
the Radiation Protection Supervisor (RPS) about the allegation and about receipt of radioactive material at the Comanche Peak Steam Electric Station (CPSES), and also reviewed their radioactive source log. The RPS stated,. i and the radioactive source log documented, that first receipt of radioac-l tive material at CPSES was on January 10, 1980, and that the quantity of i ! strontium 90 and tritium received was exempt from licensing requirements. 4 These small quantities were for laboratory use. The RPS also stated that although a leak-test procedure utilizing tritium could be employed for the j turbine generato'r unit, the turbine generator at CPSES was hydrotested j during hot functional testing, in lieu of the tritium leak-test procedure 1 outlined in the manufacturer's instruction manual. 1 The RPS stated that the first controlled shipment of tritium at CPSES was received on January 31, 1983. This shipment was authorized urder a state ! i of Texas radioactive material license (No. 5-2892) issued October 9,1980, and the CPSES radioactive source log documented this receipt. The RPS also stated that an aliquot of this tritium standard was used to prepare a j standard to calibrate the tritium monitors located on the turbine generator unit. The aliquot of tritium used to prepare the calibration standard was recovered,.and CPSES verified its original radioactivity. The RPS stated that when-the plant returns to hot functional testing after fuel load, the 4 l primary coolant of the turbine generator will be " spiked" with this tritium. During operation of the turbine generator, a small amount of hydrogen gas will be extracted and measured by a tritium monitor. Thus, if there was a leak within the turbine generator cooling system, tritium would be detected in the hydrogen gas. Y K-105 i
The TRT reviewed a Texas Utilities Generating Company (TUGCO) startup test log, which documented that the turbine generator primary coolant system and components were pressure-tested, not tritium-tested, in accordance with test procedure CPM 6.9I, " Main Generator Primary Water and Seal Oil." These tests began on December 4,1982, and concluded on October 11, 1983. The TRT reviewed the CPSES-established documented program for controlling radioactive source material, as outlined in health physics administrative procedure (HPA)-105; for receipt of radioactive material, as outlined in health physics instruction (HPI)-202; and, for shipment of radioactive materials as outlined in HPI-203. The TRT also learned that the Region IV staff conducted extensive reviews of these programs as part of the pre-operation inspection program, and documented the results of these inspec-tions in Region IV Inspection Reports 50-445/83-16, 83-35, 84-02, and 84-25. The TRT found no evidence to support the mishandling allegation.
- 5. Conclusion and Staff Positions: Based on reviews of CPSES procedures, radioactive source log sheets, startup test logs cod startup test data sheets, and on interviews with the RPE and the RPS, the TRT concludes that CPSES received radioactive material (which was exempt from NRC licensing requirements) approximately in January 1980, but did not receive a licensed shipment of tritium prior to January 1983. The NRC staff found no evidence to support the allegation that radioactive material was " dumped" into the Comanche Peak reservoir at CPSES during the period from September 1978 to November 1978.
The TRT tvied to contact the alleger during the inspection; however, the alleger us unavailable. Therefore, the TRT was unable to provide the above findings and conclusion to the alleger. A letter was sent to the alleger on January 22, 1985, requesting that he participate in an inter-view with the TRT; however, there has been no response to this letter.
- 6. Actions Required: None.
K-106
1 i I. .
- 1. Allegation Category: Miscellaneous 5, High Pressure Turbine l 2. Allegation Number: AM-6 :
- 3. Characterization: It is alleged that cracks were observed in the lower i' casing of the high pressure (HP) turbine. The alleger did not specify '
the reactor unit where the cracks were alleged to exist.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT).
! could not contact the alleger because the alleger's telephone number and address were unknown and relatives would not provide the TRT with this information. ;
1 1 The TRT reviewed the applicable design, procurement, and vendor inspection
- information described in the Final Safety Analysis Report (FSAR), Sec-tion 10.2, " Turbine Generator"; Allis-Chalmers, Turbine Description, j No. 1-1-0200-7163; TUGC0 Purchase Order, No. CP-0003; and, Vendor Surveil-
, lance Report No. 294. The HP turbine was supplied by Allis-Chalmers Power Systems, Inc., and is classified as nonsafety related. The HP casings were fabricated in Japan by a casting process; however, they were subsequently i shipped to the Mulheim plant in Germany where manufacturing was completed. l The vendor surveillance referenced above was performed in Germany by a
; Gibbs & Hill (G&H) QC engineer on October 3-6, 1977. Records indicate that the engineer visually examined the Unit 1 HP turbine, wit = eed '
hydrostatic testing, and reviewed the ' supporting documentatios., .ch included nondestructive examination (NDE) records. The vendor surveillance report on the Urit 1 HP turbine concluded that inspection and testing showed the casing to be satisfactory.
- On August 4, 1984, the TRT interviewed both the Texas Utilities Electric l Company (TVEC) supervisor of turbine ~ construction and the TUEC lead startup i engineer. They stated that they observed no cracks in the casing. In
! addition, the lead startup engineer was present when the Unit 1 turbine l was rolled to synchronous speed during testing, and he indicated that no casing problems (including casing leaks) were observed and that only an j insignificant diaphram leak was detected during testing. The TRT also reviewed test documentation which showed acceptable results.
The TRT inspected the outside of the upper and lower casing of-the HP tur-4 bine for Unit 2 and found no cracks. They did not inspect the casing of the Unit 1 HP turbine because it is nonsafety related and is wrapped with j insulation. Since independent vendor inspection and test records, and TUEC observation and testing, revealed no unacceptable conditions, the TRT l did not request removal of the insulation or did not perform further casing
- inspections of the Unit 1 turbine.
- 5. Conclusion and Staff Positions: The TRT determined that TUEC and vendor personnel performed visual inspections, witnessed hydrostatic and startup testing on the HP turbine for Unit 1, and found no unacceptable conditions.
! The TRT also inspected the outer casing of the Unit 2 turbine and found no l cracks. Moreover, this equipment is classified as nonsafety related and is'not needed for.the safe shutdown of the reactor. l i i , K-107 1
-,,_m.,-- - . . , .-,-r..,..-s.-, - - . ~ . - , , ~ . - ,- - . . - . . - , , ..,--. - , - -,,.,.,-. - ,-,,, .,- ,- c-, - , , , - . , ,
The TRT attempted to contact the alleger during the TRT inspection to obtain specifics, but the alleger declined to be interviewed. Consequently, the TRT was unable to present.its findings and conclusions to the a11eger in a follow-up interview. A letter was sent to the alleger on January 22, 1985, requesting that he recondiser participating in an interview, with the TRT; however, there has been no response to this letter.
- 6. Actions Required: None. .
3 4 l l l S K-108
- 1. Allegation Category: Miscellaneous No. 6, Pressurizer Area Piping
- 2. Allegation Number: AM-7
- 3. Characterization: It is alleged that an 18-inch section was cut from a prefabricated pipe "in the pressurizer area."
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) attempted to contact the alleger to determine the exact location of the cut piping because it was not specifically identified in the allegation.
However, the alleger's telephone number and address were unknown, and rela-tives of the alleger would not provide the NRC staff with this information. Due to the size of the piping (18 inches), and because the alleger stated that the " cut pipe was close to the pressurizer," the staff assumed that the cut section of piping was located in the pressurizer system. Based on a review of a Final Safety Analysis Report (FSAR) flow diagram (" Reactor Coolant System," Section 5.1) and a walkdown inspection of the pressurizer area, the NRC staff determined that there is no 18-inch line in the pres-surizer system and immediate area and that the 12-inch piping that runs from near the top of the pressurizer (at approximately the 907-foot eleva- ! tion) to the pressurizer relief tank (at the 834-foot elevation) most closely fit the description given in the allegation. The TRT also reviewed the isometric drawings for Unit 1 and Unit 2. These show the as-built configuration of the four 6-inch lines which run from the pressurizer to a common 12-inch line which, in turn, runs to the pres-surizer relief tank. In this run of piping are both smaller piping and a 3-inch line which runs from the 12-inch line to the safety relief valves, and which ties into piping in the residual heat removal system (RHR) suc-tion (Trains A and B) and for chemical and volume control system (CVCS) seal return and letdown. All modifications to the pipe lines identified in the isometric drawings, including the trimming of pipe for fitup, were recorded on component modification cards (CMCs). Trimming modifications of 7/8-inch and 1-1/2-inch made on 12-inch piping were recorded on CMCs 61551 and 47943R1. The TRT inspected and counted the number of welds in r.ll of the above piping runs, then compared that number with both the num-Der shown on the as-built isometric drawings and the number on the CMCs. All three numbers were the same, and the NRC staff found no evidence of unauthorized work in the piping system. Because the 12-inch piping runs are nonsafety related and not essential for safe shutdown of the plant, the quality assurance requirements of 10 CFR Part 50 Appendix B are not applicable; however, Texas Utilities Electric Company (TUEC) technicians monitored the piping installation in accordance with CP-CPM-6.9, " Welding and Related Processes." The TRT selected and reviewed two welding records and two test reports which confirmed that appropriate procedures were used to control welding. Again, the TRT found no evidence of unauthorized work. K-109 l
' TUEC nerformed a successful final system check from the pressurizer to the pressurizer safety relief valves, including the piping down through the The system RHR suction and the CVCS seal return letdown relief valves. check was made concurrently with hydrostatic tests IRC-101 and IRC-01A on August 31 and October 19, 1982. The actual test pressure was 1113 psig for 21 minutes, which meets the ANSI B31-1 code requirement which is 1.5 times the design pressure of 700 psig or 1050 psig. The diameter of the Class 1 ASME pressurizer piping did not fit the description contained in the allegation. However, the TRT inspected the Class 1 piping from the pressurizer to the upstream side of the pressurizer safety relief valves and reviewed the as-built isometric drawings, noncon-formance reports (NCRs), CMCs, and N-5 ASME data forms for the 3-inch and 6-inch piping to ensure that modifications or changes had been authorized and were recorded. Based on this review, the TRT found no evidence of unauthorized work and determined that the QA records for ASME piping were in order. In addition, the NRC staff inspected the pipelines visible at 4 floor level, and the number of welds appeared to correlate with the as-built isometric drawings and with the information on the N-5 data forms required by ASME Code.
- 5. Conclusion and Staff Positions: Based on inspections of the piping and on a review of applicable documents, the TRT found no evidence which would support the allegation that unauthorized cuts or welds were made in piping from the pressurizer to the pressurizer relief tank, the RHR suction relief valves in Trains A and B, the CVCS seal return, or the CVCS letdown piping systems close to or in the pressurizer area. Accordingly, this allegation has neither safety significance nor generic implications.
The TRT was unable to provide its findings and conclusions to the alleger, because the alleger declined to be interviewed. A letter was sent to the alleger on January 22, 1985, requesting that he recondiser participating in an interview, with the TRT; however, there has been no response to this letter.
- 6. Actions Required: None.
) 110
l
- 1. Allegation Category: Miscellaneous 7, Condenser
- 2. Allegation Number: AM-8, AM-9 and AM-10
- 3. Characterization: It is alleged that: (a) the Unit 1 main condenser tubes were beaten with air and sledge hammers, were split during belling and flaring, and were improperly rolled; (b) the wrong type condenser for the steam generator was used; (c) the tube support sheets had holes that were misaligned by 3/8 inch; and, (d) the turbine-to-condenser tubing was mis-aligned and jacked into alignment, causing stress.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) interviewed the alleger on August 24, 1984, to gather additional infor-mation regarding his concerns.
A Region IV letter to TUEC identified the potential safety concerns asso-ciated with this allegation and asked Texas Utilities Electric Company (TVEC) to respond in writing. In a letter dated July 9, 1984, TUEC out-lined their monitoring procedures for condensers during and after con-struction activities, as well as the various tests that they had conducted. This letter stated that testing demonstrated that no leakage of lake water into the steam generator occurred because of differential pressure that is maintained during operation. It also states that the main condenser and circulating water systems are not s4Tety related and therefore r:ot sub-ject to the quality assurance requirement of 10 CFR Part 50, Appendix B. Regardless, construction and fabrication were done in accordance with sound engineering and construction practices. In assessing this allegation, the TRT reviewed NRC Office of Investigations interview 84-006; the RIV letter and TUEC response referenced above; TUEC office memoranda SU-81051, 50-81081, and SU-81134; and Condenser Retest No. 1 CP-AT-27-01. These documents show that TUEC was well aware of and was controlling and correcting problems associated with the condenser fabrication and operation. The TRT attempted to visually inspect the condenser internals, but this was not possible because of the congestion caused by the tube bundles, the internal piping, and the bracing. Therefore, the inspection was limited to using a hand-t.eid light and observing the bundles from the manway opening. Observation was difficult because of the distance between the manway and the tube bundles and because of surface rust; however, the limited inspection detected no irregularities. The TRT interviewed three employees who had direct knowledge of the work on the main condenser tubes. With respect to part (a) of this allegation, i.e., use of hammers and splitting tubes, a Brown & Root (B&R) Millwright Superintendent stated that hammers were used on the condenser tubes, but only after the installation of a special tool to pro *.ect the tube ends. He also stated that this is a standard practice in condenser assembly. Concerning part (c) of this allegation, the Millwright Superintendent disclaimed any knowledge of misaligned tube sheet holes. The TRT also interviewed a TUEC Operations Results Engineering Supervisor concerning the availability of documents such as inspection reports, l K-111 1 I
deficiency reports, in progress monitoring reports, nondestructive examina-tion (NDE) reports, and procedures related to the allegation. He stated that work was performed to B&R procedures and that he did not keep any official records because there was no requirement; therefore his particular involvement was limited to nonsafety-related surveillance for commercial consideration. When asked about tube-rolling problems, he stated that the variance in manufacturing tolerances of the tube outside diameters and the tube sheet hole diameters did present a problem, part (c), in the beginning, but the problem, part (a) of this allegation, was solved when they properly calibrated the rolling tool. He also stated that as the work progressed, his level of confidence in the craft personnel's work reached .a point where 100 percent surveillance was not required. He added that during the rolling of the' tubes, the optimum tube-wall reduction for a positive tube-to-sheet seal was in the range of 6 to 9 percent and that in the beginning, when experimenting with the torque on the rolling tool, although the 9 percent i reduction in wall thickness did occur in some cases, it did not result in any of the tubes cracking. In another interview, a B&R Millwright Foreman informed the TRT that he was not involved in the construction of the condenser and disclaimed any knowledge of the alleged problems. Part (b) of this allegation was dis-i cussed with the Foreman, who said it was rumored that the condensers, the auxiliary condensers, and the moisture separators might be retubed. This rumor was confirmed both by the Millwright Superintendent and in a tele-phone conversation with the TUEC Nuclear Engineering Manager. The latter informed the TRT that his group was commissioned to make a feasibility study of condenser etubing, first doing Unit 2 and then doing Unit 1 during the first refueling outage. The proposed change was to retube using titanium tubes instead of chrome-nickle (Cr-Ni) tubes in order to I raise the pH level of the water on the secondary side from a level of 9.4 to between 9.8 and 10.0. The alleger may have been referring to this pro-posed change when he said "it's the wrong type condenser for the type of steam generator." However, Westinghouse provided Gibbs & Hill, Inc. (G&H) with Specification 2323-MS-23, which contained the original design, oper-ating conditions and criteria that were to be met, and this specification was used. The second design of the tubing was proposed as a design improvement. On September 14, 1984, the TRT contacted an Allis-Chalmers (A-C) represen-tative who supplied information regarding part (d) of this allegation. Installation progress reports showed that the low pressure (LP) II con-denser's expansion-joint weld was completed on September 18, 1978. The welding process was monitored during welding, and once in approximately every hour micrometer readings were taken at 12 points around the joint. These readings showed that the maximum movement was 0.026 inches (0.660 mm), which was acceptable to Allis Chalmers, Westinghouse, and TUEC. Other than visual inspection, an NDE was not required. The same progress report showed that the LP-I expansion joint welding began on the same date. On October 13, 1978, deviation report No. 14094 was written against the lower weld on this joint. The movement of 0.277 inches (7.030 mm) exceeded that allowed by A-C. Consequently, on October 19, 1978, the removal of the lower weld was begun, and some 1,500' inches of 3/4-inch weld were removed. The micrometer logs kept by K-112
I the TUEC Operations Maintenance Foreman showed that rewelding began on November 13, 1978, and was completed on November 17, 1978, and that move-ment was controlled to a maximum of 0.023 inches (0.584 mm). All welding was done to B&R weld procedure (No. 10046) and to Westinghouse directions for skip (intermittent) welding. Considering these welding controls, there was no evidence that jacking occurred or that any undue or undesir-able stress was introduced into the welded joints. Following welding, the only NDEs performed (other than visual inspections) were hydrostatic and vacuum tests which were successfully completed. The TRT also reviewed seven condenser hydrostatic test packages for the results of tests conducted in accordance with the G&H test procedure (Specification 2323-M-23). These tests spanned the period from December 1980 to April 1984. The TRT found that retests of the shell were made following design changes. The last of these retests, No. ICO-0200E, was performed in September 1983, and involved sodium tracer injection and sampling at points through the condenser wall. In all cases the tests were successfully completed. In May 1984, TUEC performed a vacuum and water box priming retest (Procedure ICP-AT-27-01-RT-1) to again verify that the main condenser could be evacuated by the vacuum pumps and hold vacuum for 1 hour. The system test engineer witnessed and accepted this sucessfully completed test. The lead startup engineer, the manager of plant operations, the TUSI nuclear engineering manager, and the manager of nuclear operations also reviewed and accepted the test results. The TRT's review of these documents indicated that there was no evidence of an over-stress problem with either the expansion joint or the piping connection.
- 5. Conclusion and Staff Positions: The TRT determined that this unit is classified as nonsafety related and is not essential for the safe shutdown of the plant, and confirmed that when TUEC began the tube-rolling procedure, they experienced some fabrication problems; however, these problems appear to have been solved. The TRT also confirmed that TUEC plans to retube the condenser with titanium tubes to improve its design and operation.
The TRT found no evidence that holes in the tube support sheets were mis-aligned by 3/8-inch, nor did they find evidence that the turbine to con-denser was misaligned to the point that excessive stress was introduced. The conclusive evidence is that the condenser was constructed following approved construction and testing procedures and, as such, will perform its design function. The TRT concludes that this allegation was not substantiated. However, it was true that fabrication problems occurred and that condenser redesign-(tube material changes) and misalignment occurred, but not as alleged. Accordingly, this allegation has neither safety significance nor generic implications. The TRT attempted to present its finding and conclusions to the alleger in a follow-up interview, but the alleger could not be located. The TRT was unable to find either a telephone number of an address for the alleger. A letter will be sent to the intervenor of record, outlining the resolu-tion of the alleger's concerns.
- 6. Actions Required: None.
K-113
1 l
- 1. Allegation Category: Miscellaneous 8, Damaged Component Cooling Water Tank Supports
- 2. Allegation Number: AM-12
- 3. Characterization: It is alleged that during the installation of Unit 1 component cooling water (CCW) surge tank, the anchor bolts were damaged.
- 4. Assessment of Safety Significance: The NRC Technical Review Team inter-viewed the alleger on August 24, 1984, to gather additional information regarding his concerns.
The TRT reviewed the background material which alleged that the anchor bolts were beaten sideways with a hammer to make them line up with holes in the plate and were overtorqued to the point that they stretched. The CCW surge tank, part of the component cooling water system, is listed in Final Safety Analysis Report (FSAR) Table 17A-1 as Safety Class 3, ASME III
- Code Class 3, and Seismic Category I. The centerline of the CCW surge tank for Unit 1 is at an elevation of 889 feet, 6 inches in the Auxiliary Build-ing; the baseplate attaches to bolts embedded in concrete which interface at an elevation of 895 feet, 6 inches.
In assessing this allegation, the TRT reviewed Texas Utilities Services Inc. (TUSI) Drawing N-2640-359. The TRT also visually inspected the installed CCW surge tank and saw no stripped threads or bent or cracked bolts. Two nuts and one washer were present on each bolt. The CCW surge tank has 10 bolts on each end which support the tank. A review of the installation documentation revealed that 5 out of the 20 concrete anchor bolts were misaligned. A TUSI letter (CPP-00825), dated March 2, 1979, documents the need for modifying the baseplate holes. This letter, which de. scribes the misalignment, indicates that the bolts were not installed as required by the specification and drawing. As a result, component modification card (CMC) No. 4263 was approved June 8, 1979. The TRT found no nonconformance reports (NCRs) in the quality assurance (QA) records vault. The misalignment may have caused installation problems; however, if any damage occurred at that time it was not documented. Traveler No. ME78-108-1101 and a traveler revision record sheet document the installation of the CCW surge tank. This traveler was initiated on October 10, 1978, and the tank was placed on June 13, 1979, when the Millwright Supervisor signed the traveler. This traveler failed to give instructions for tightening nuts on anchor bolts as required by Procedure 35-1195-MCP-1, Revision 2, paragraphs 4.1.10 and 4.1.11. Both the quality control (QC) inspector and the Millwright Supervisor signed the traveler on June 18, 1979, indicating that the tank was level and located as recorded on the as-built drawing. The TRT' interviewed the Millwright Supervisor and the QA/QC and engineer-ing personnel listed on the traveler and learned that although they had direct knowledge of work activities on a day-to-day basis, they had no knowledge of bent or cracked bolts or of damaged threads. The engineer K-115
stated that the millwrights routinely brought any problems to him, and he could not believe craftsmen assigned to him would carelessly bend and damage the bolts. When reviewing this allegation, the TRT found that design change authori-zation (DCA) No. 9909, Revision 1, dated April 10, 1981, documented modi-fications made to the structural support saddles to increase their strength and to meet seismic requirements. Travelers No. CE-82-143-1100 and No. ME-81-1563-2-1101 documented that this work was done on Units 1 and 2. DCA No. 11468, Revision 10, dated May 1, 1984, documented addi-tional seismic brace work on Units 1 and 2 tank supports. In their review of the applicable documentation, the TRT found no evidence indicating problems with the installation of the Unit 2 CCW surge tank or with misaligned or damaged bolts. The TRT discussed the load on the anchor bolts used to install each of the Unit 1 and 2 tanks with the cognizant Gibbs & Hill Inc. (G&H) engineer, who provided and explained the data in G&H Calculation Number SAB-104, Set
- 4. The full tank load was calculated to be 51.5 ksi (1000 lb/ square inch).
Based on these values, the worst case analysis for a seismic event was
' calculated to determine the tensile and shear loads on the 1-inch anchor bolts (ASME A320-A7, 105,000 ksi yield strength). These bolts and the bracing and supports, added as a result of modifications, were documented as being strong enough to carry the loads and meet safe shutdown earthquake requirements. Nuts used to bolt the west end of the tank to the concrete beam are required only to be hand-tight with a locknut; therefore, no torquing was required. On the east end, torquing and a locknut are required. ,
- 5. Conclusion and Staff Positions: Based on its review of applicable documen-tation and interviews with cognizant personnel, the TRT concludes that problems were experienced during installation of the Unit 1 CCW surge tank because of misaligned bolts and that the necessary modifications were made after engineering review and approval. The TRT found no evidence to support the allegation that these bolts were beaten with hammers and were overtorqued to the point of stretching and cracking them. Accordingly, this allegation has neither safety significance nor generic implications.
The TRT tried to provide the above findings and conclusions to the alleger in a follow-up interview, but the alleger could not be located. The TRT was unable to find either a telephone number or an address for the alleger. A letter will be sent to the intervenor of record, outlining the resolu-tion of the alleger's concerns.
- 6. Actions Required: None. However, if a violation is issued TUEC will be required to take corrective action and respond.
K-116
- 1. Allegation Category: Miscellaneous 9, Hayward Tyler Pump Deficiencies
- 2. Allegation Number: AM-13
- 3. Characterization: It is alleged that Comanche Peak Steam Electric Station (CPSES) has pumps in safety systems manufactured by Hayward Tyler Pump Co.
(HTPC) that may have unidentified deficiencies because of a poor quality assurance (QA) program by HTPC.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) interviewed the alleger on August 24, 1984, to gather additional infor-mation regarding his concerns.
The NRC Technical Review Team (TRT) learned that in late 1981, the NRC received allegations that upper management of Hayward Tyler Pump Company of Burlington, Vermont failed to support the quality assurance program. In 1982, the NRC Region IV staff inspected the HTPC QA program after receiving these allegations. The investigation established that signifi-cant deficiencies existed in the implementation of HTPC's QA program from 1977 through 1981. As a result, the NRC issued a report and a Notice of Nonconformance on December 22, 1982. As a result of the above findings, the NRC also issued IE Bulletin No. 83-05 to licensees and applicants on May 13, 1983. This bulletin addressed HTPC's failure to effectively implement its QA program, and required that specific actions be taken by the holders of operating licenses and construction permits who were using or planning to use HTPC pumps in safety systems. IE Bulletin 83-05 requirements for successful pump operability applied to both the original pump assembly and to reassembled pumps which use spare parts. The bulletin required the following actions from NRC applicants and licensees: to provide NRC with the number of HTPC pumps and their service appli-cation, to provide NRC with a summary of the in-service test requirements for the affected pumps and spare parts, to conduct a pump performance test by running the pump continuously for a minimum of 48 hours without maintenance or repair, with the test incorporating specific criteria provided by HTPC, to provide NRC with the results of the required ASME Code hydrostatic pressure test, to implement HTPC recommendations with respect to fitup and dimensional considerations during installation of spare parts, to conduct a pump performance test when spare parts are installed, unless it could be demonstrated that the spare part in question would not affect any parameters that are measured, and K-117
to provide NRC with the results of the required ASME Code hydrostatic pressure test on spare parts that form part of the ASME Code pressure boundary. On May 24, 1982, an NRC Region IV inspector testified before the Atomic Safety and Licensing Board (ASLB). On page 99 (Answer 64) the RIV inspec-ter stated that Hayward Tyler pumps were used at CPSES and allegations or problems were being investigated. This was preceded by an allegation made by an unknown alleger in March 1982, and may have led to the'ASLB questions about these pumps. The TRT determined that an NRC inspector had reviewed the Texas Utilities Electric Company (TVEC) letter dated August 10, 1983, which was in response to IEB 83-05, including the supporting documentation, to assure that all required actions were addressed and had been performed. The inspector had determined that the use of HTPC pumps at CPSES is limited to the station service water systems and all required actions had been docu-mented as complete with respect to the two CPSES Unit 1 station service water pumps (SSWP). The 48-hour endurance test requirements were met and exceeded when the Unit 1 SSWPs were operated continuously between March 3 and May 26, 1983. The number of hours accumulated during that run totaled 1528.75. The NRC inspector verified this by a review of the CPSES Unit 1 Control Room reactor operator log. TUEC's supervisor of technical support and startup engineering stated that the SSWPs had accumulated approximately 16,000 operational hours since initial startup in February / March 1981, without any major repairs or any significant degradation in performance. The two Unit 2 SSWPs will be tested during the Unit 2 preoperational testing program, which incorporates the requirements of IEB 83-05.
- 5. Conclusions and Staff Position: The TRT concludes that TUEC had identified Hayward Tyler pumps onsite and tested the pumps and reported as required by IEB 83-05. The TRT also concludes that the allegation had potential safety significance and generic implications; however, TUGCO's compliance with IEB 83-05 has eliminated those concerns with respect to CPSES Unit 1 SSWPs. The Unit 2 pumps will be inspected under Unit 2 preoperational testing.
The TRT attempted to provide the findings and conclusions described above to the alleger, but the individual could not be located. The TRT was unable to find either the telephone number or an address for the alleger. A letter will be sent to the intervenor of record, outlining the resolu-tion of the alleger's concerns.
- 6. Actions Required: TUEC shall verify compliance with IEB 83-05 require-ments for CPSES Unit 2 SSWPs during preoperational testing for Unit 2.
1 K-118
i I
- 1. Allegation Category: Miscellaneous 10, Damaged Diesel Generators
- 2. Allegation Number: AM-14
- 3. Characterization: It is alleged that the Unit 1, Train A, diesel generator was damaged in May 1982, because of improper handling practices.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) tried to contact the alleger during its inspection to learn more details about the allegation. The TRT reviewed all nonconformance reports (NCRs) issued in May 1982, that pertained to the emergency diesel generators (EDGs). Of the 11 NCRs reviewed, 4 (E-82-005335, E-82-00560, E-82-006065, and E-82-004795) documented equipment or instrument damage; however, the necessary corrective action was taken and the NCRs were closed appropriately.
1 The TRT determined that it was unlikely that any damage that occurred in May 1982, could now affect the operability of the EDGs in light of the extensive EDG inspection and testing that took place in 1984. This inspec-tion and testing was the result of generic EDG problems which Transamerica Delaval, Inc. (TDI) and owners of TDI's EDGs identified to the NRC. On August 12, 1983, the main crankshaft on one of the three EDGs at Shore-ham Nuclear Power Station broke into two pieces during a load test. TDI issued several 10 CFR Part 21 reports that reflected a variety of major and minor defects. These defects included cracks in piston skirts, push rod cracks, governor drive coupling failures, potential failures in fuel lines, and dimensional problems with component fasteners and dowel pins. Although there are some design differences between the EDGs at Comanche Peak Steam Electric Station (CPSES) and those at other plants, the identified defects were generic in nature. During the evaluation of the failure and repairs of the Shoreham EDGs, information related to the operating history of TDI engines and the QA program of the manufacturer was identified which called into question the . reliability of all TDI diesels. As a result of the evaluation and its generic implications, representatives from affected nuclear power plants formed an " Owners' Group" to investigate all aspects of the quality and reliability of TDI supplied EDGs. The Owners' Group developed a generic inspection program. This program addressed the specific concerns brought about by defects reported to the NRC by owners of (DI EDGs, 10 CFR Part 21 reports from TDI, and other areas of concern in order to develop adequate confidence in these EDGs. Texas Utilities Electric Company (TUEC) also expanded some inspections of the TDI EDGs by increasing sample sizes, inspecting other areas on their own initiative, and inspecting both of the Unit 1 EDGs between February and June 1984. The inspections included disassembly and nondestructive examinations of parts using methods such as radiography, liquid penetrant testing, magnetic particle testing, visual inspections and measurements, eddy current testing, ultrasonic testing, and metal comparator testing. TUEC then transmitted the results of their inspections to the Owners' Group for evaluation and incorporation into the recertification process. K-119
The EDGs were reassembled after cleaning, and the inspections and non-destructive tests were completed. This effort was closely controlled by approved procedures and by QC surveillance. TUEC replaced parts which had been identified as containing potential generic defects and also replaced those parts found to have defects previously unidentified. Upon completion of assembly, TUEC retested each EDG by performing the ' entire portion of the preoperational test program which involved operation of the EDGs. The TRT determined that TUEC made the following tests and that the test results were satisfactory. 1CP-PT-29-01, RF1 " Diesel Generator Auxiliary Systems, Retest 1" 1CP-PT-29-02 " Diesel Generator Control Circuit Functional and Start Test" 1CP-PT-29-03 " Diesel Generator Load Tests" 1CP-PT-29-04 " Diesel Generator Sequencing and Operational Stability Test" 1CP-PT-29-05 " Diesel Generator Reliability Tests" The TRT learned that the NRC Region IV (RIV) Resident Inspector for Operations conducted inspections on nearly a daily basis, starting with the disassembly process in February 1984, and ending with the witnessing of the testing in August 1984. This NRC inspection effort included (but was not limited to) observation of the work and testing in progress, review of procedures used and compliance ther'eto, and tracking the work to ensure that TUEC followed the Owners' Group program and adequately documented results. NRC Inspection Reports 50-445/84-07, -15, -17, -18, and -20, which document this inspection effort,' indicate that the recertification program was satisfactorily completed. The TRT determined that extensive engineering evalutions and tests had been conducted by TDI, the Owners' Group, and TUEC. The RIV inspectors reviewed and witnessed the satisfactory testing of the Unit 1 EDG. Therefore, if any damage did occur, it had been corrected before preoperational testing.
- 5. Conclusion and Staff Positions: The TRT found documented evidence support-ing the alleger's concerns about damage to'the EDGs in the four NCRs listed above and concluded that this allegation had potential safety significance and generic implications. However, since appropriate corrective action was taken and documented, the TRT concludes that this damage no longer exists.
In addition, any damage affecting the reliability and operation of the EDGs that was not documented would have been discovered and corrected during a comprehensive recertification program undertaken by TUEC. The TRT review of TUEC and NRC Region IV documents indicates satisfactory completion of the above retests. The TRT review also confirms that the EDGs will perform in accordance with design. Accordingly, the TRT finds that this allegation no longer has either safety significance or generic implications. The TRT interviewed the alleger and provided the above findings and con-clusions. The alleger indicated that'his concerns were resolved. ;
- 6. Actions Required: None.
K-120 .
- 1. Allegation Category: Miscellaneous 11, Polar Crane Shimming
- 2. Allegation Number: AM-15, AM-16
- 3. Characterization: It is alleged that the shimming of the Unit 1 polar crane rail system supports was improper and that the polar crane system is improperly installed.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) tried to contact the alleger during its inspection to learn more details about the allegation. On August 14, 1984, two TRT members visually -
inspected the shims from the polar crane. During the first 180 rctation of the crane, the TRT members stood on the platform above the operator's booth to view the radial restraint brackets and the seismic restraint brackets. Several brackets at different locations appeared to have gaps in' excess of 1/16 inch. However, this only confirmed what had been previously observed by an NRC Region IV Resident Inspector and was docu-mented in NRC Inspection Report (IR) No. 50-445/84-08, which required corrective action which was not yet completed. The TRT then moved below the operator's booth to view the polar crane rail system from another vantage point. The TRT observed the shims used for shimming 28 crane girder to girder-support brackets. During this 180 rotation, the TRT observed large gaps, particularly on the inside edge (looking from the inside of the Containment Building to the outside). The TRT met with the Texas Utilities Electric Company (TVEC) project civil engineer, the Brown & Root (B&R) project control manager, the B&R subcon-tracts supervisor, and a representative from Chicago Bridge and Iron (CB&I) to determine the gap-tolerance specification between bearing plate "A" (Dwg. 2323-S1-0515, Revision 4) and the girder to girder-support bracket. Neither Gibbs & Hill (G&H) specification SS-14 nor the Crane Manufacturers Association of America Manual (CMAA-70) addressed this issue. The meeting failed to produce a specific answer; however, copies of two letters related to the issues were provided. The first, a B&R letter (No. BRF-7404), dated November 8, 1977, contained the as-built measure-ments of gaps at all shim locations and a request for G&H to evaluate this information and provide direction. At 28 locations, the as-built drawings showed gaps that ranged up to 0.581 inch. In the second letter, G&H (GHF-2207, dated November 28, 1977) responded as follows: Girder Seat Connections These seated connections will not require shimming since the area in bearing is at least the width of the bottom flange of the crane girder. The gap dimensions indicated in the Brown & Root survey exist only at the extreme edges of plate A, Section 3-3, Dwg. 2323-51-0515, Revi-
- sion 4.
The TRT noted that the bottom flange of the girder referenced in the G&H letter (the bearing surface) is 20 inches wide. On August 30, 1984, an NRC inspector, accompanied by a TUEC quality con-trol (QC) inspector, inspected the 28 crane girder to girder-support K-121
.-__x -.- . - . - . . _- - . -- - , -, _ . -
bracket shims. Nine girders, identified as A7-6 right-end (RE), A7-8 (RE), A7-12 (RE), A7-14 (RE), A7-18 left-end (LE), A7-19 (RE), A7-20 (RE), A7-24 (RE), and A7-25 (RE), had gaps in excess of 1/16-inch extending under the bottom flange. This observation invalidated the G&H assumption of 20 in-ches of bearing surface. The TRT closely observed girder A7-20 (RE) as the crane wheels passed directly over the support bracket and saw no visible compression (closure)
~
of the gap. In addition, a visual inspection of the complete rail system revealed that the rail has moved or is meving circumferential1y, as indi-cated by the fact that some of the 1-inch-diameter stabilizing rods are bent from the force of this movement. The 3/8-inch designed gap between the ends of the rail section also varied from 0.000-inch to 0.875-inch, when measured at the inside edge of the rail. In addition, three of the rail-to-rail ground wires and two Cadwelds were broken, and at least two rail shim plates had partially worked out from under the rail. The TRT interviewed the polar crane operator and asked if he knew of any existing problems with the crane or its operation. He replied that the crane operates satisfactorily and has experienced no apparent problems. He also stated there are no " dead spots" (i.e., no loss of electrical energy at spots) in the bus bars. The TRT found additional shimming problems and additional types of problems, described above, that had not previously been identified. These deficien-cies appear to be safety significant and generic.
- 5. Conclusion and Staff Positions: Based on the above inspections, the TRT concludes that this allegation is substantiated and is potentially safety significant. The problem with shimming and inspection of safety-related work was first identified in 1982. Because problems still existed in 1984, this matter appears to be generic.
On November 8, 1984, the TRT interviewed the alleger to provide the above findings and conclusions. The alleger stated that his concerns were resolved.
- 6. Actions Required: TUEC shall inspect the polar crane rail girder seat connections for the presence of' gaps which reduce the bearing surface to less than the width of the bottom flange.
TUEC shall perform an analysis which will determine whether existing gaps are acceptable or if corrective actions are required. TUEC shall determine if additional rail movement is occurring and, if so, provide an evaluation of safety significance and the need for corrective action. TUEC shall perform a general inspection of the polar crane rail and the rail support system, correct identified deficiencies of safety signifi-cance, and provide an assessment of the adequacy of existing maintenance and/or surveillance programs. I K-122 f l
l l Note: The gaps in the seismic restraints were the subject of NRC Inspec-tion Reports 50-445/82-11, 50-446/82-10, and 50-445/84-08; violations were issued in each report. Although these matters may have been evaluated and a response made to the referenced violations, TUEC shall consider this matter as a part of the inspection of the polar crane system. 6 1 4 T ), , K-123 i i
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- 1. Allegatian Category Miscellaneous 12, Welding of Lifting Lugs onto Tornado Missile Barrier Doors
- 2. Allegation Number: AM-17
- 3. Characterization: It is alleged that deficient welds on a missile barrier door were accepted.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) tried unsuccessfully to contact the alleger during its inspection to learn more details about the allegation. The TRT identified the location of the doors from information provided by Texas Utilities Electric Company (TVEC) quality assurance personnel who were present at the time of the allegation, and from a review of the Atomic Safety and Licensing Board (ASLB) hearing record on intimidation and harassment. The door referred to in this alle-gation is the tornado missile barrier located at ground level on the west side of the Unit 1 diesel generator room. The alleged deficient welds are 18 double grooved welds that attach the lifting lugs to the three missile barriers. These welds were terminated by wrapping the weld around the end of the lug, a practice questioned by the alleger.
The TRT learned that the alleger mistakenly believed that neither the welds which were made nor the wraparounds terminating the welds were allowed by the weld symbol specified on the drawing. The alleger thought that a lifting lug which was welded to the flat side of the missle barrier steel door (using a double grove T-weld joint) should have been indicated with a weld symbol showing a double groove on each side of the lifting lug and with a weld symbol at the end of the lug showing a fillet weld where the runoff occurred. The Brown & Root (B&R) inspectors who actually performed the inspections did not interpret it as the alleger did and accepted the 18 disputed welds. The TRT reviewed the B&R inspection reports, which indicated that welds were performed in accordance with Welding Procedure Specification (WPS) 10046, Rev. 9, and with American Welding Society (AWS) code requirements. The TRT also reviewed the construction traveler and the inspection reports for the shop fabrication and field-fitting of the missile barriers and found that, although rework occurred, no rework was done on the lifting lug welds. The TRT reviewed the inspection procedure, weld procedure, and inspection reports referred to in the traveler for applicability and compliance with AWS Code 01-1, Sections 4.6.1 and 4.6.2, to determine if the code allowed wrapping the weld around the end of the lug. These sections of the code state: 4.6.1 Groove welds shall be terminated at the ends of a joint in a manner that will ensure sound welds. Whenever possible, this shall be done by the use of extension bars or run-off plates. i i K-125
4.6.2 In building construction, extension bars or run-off plates need not be removed unless required by the Engineer. The TRT inspected the four tornado missile barriers (east and west of the diesel generator room, Units 1 and 2), except where the missile barriers had been removed in Unit 1. In Unit 2 (east side), one segment of one missile barrier was in place for trial fitting; two segments were being fabricated on location. The TRT visually inspected the lug welds on the tornado missile barrier segments and determined that the welds showed very good workmanship and that wraparound on the ends of the lugs was both minimal and acceptable. The TRT found that welding symbols had been correctly interpreted and that all of the welding described above, which included wrap around to terminate the welds, had been correctly done.
- 5. Conclusion and Staff Positions: The TRT found no errors in design or inter-pretation of weld symbols or any poor workmanship on welds of the lifting lugs on the tornado missile barriers. In addition, the TRT determined that both the welding and the inspection of doors was done in accordance with specified procedures. The lugs function only for lifting the massive mis-sile barrier doors and would have little or nothing to do with protecting safety-related equipment from missiles. Accordingly, the TRT concludes that this allegation has neither safety significance nor generic implications.
The TRT will provide the above findings and conclusions to the alleger by letter.
- 6. Actions Required: None.
l 1 K-126
- 1. Allegation Category: Miscellaneous 13, Welding of Pipe Supports in Safeguards Tunnel
- 2. Allegation Number: AM-18
- 3. Characterization: It is alleged that the tube steel used to fabricate supports by welding it to baseplates in the Unit 1 safeguards "796 yard tunnel" was cut at the wrong angle, resulting in too large a gap between the tube and baseplate.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) visually observed the entire "796 yard tunnel" to determine how many sup-ports were used, and found there were several hundred tube-steel-to-baseplate weldments/ installations. A member of the TRT also attempted to contact .
the alleger to obtain additional information to determine the approximate i location, size, and configuration of the subject tube-steel-to-baseplate weldments, because without this information indiscriminate destructive test-ing of the installed supports would be necessary in order to identify the location of the alleged gap. However, the alleger refused to communicate with the TRT. The TRT visually observed the "796 yard tunnel" to identify any condition that would show improper installation or welding on the 796-foot, 6-inch elevation, and the three, short 800-foot-elevation tunnels. Subsequently, the TRT randomly selected typical pipe supports and five hanger inspection reports (DD-1-16-025-Y33R, DD-1-16-024-Y33R, SI-1-031-041-532K, SW-1-17-716-Y33K, and AF-1-002-033-Y33K) and reviewed them to determine if the documentation of the inspections required during installation was cor-rect. The inspection reports indicated that the supports were correctly i installed in accordance with Brown & Root procedures CP-CPM-6.9E, " Pipe Fabrication and Installation," Revision 7; CP-CPM-6.9F, " Fabrication and Installation of Component Supports," Revision 0; and CP-CPM-7.1G, " Piping i Supports," Revision 0. The TRT requested that a Brown & Root inspector take copies of the inspec-tion reports for the five pipe hangers to the field and repeat those steps in the inspection which could be repeated; however, no root gaps could be inspected because all welding had been completed. Inspection included
; dimensional checks, measurement of welds, checks for proper anchoring, and j visual inspection of welds. The TRT found no major discrepancies between j the inspection reports and the field conditions. The recheck of the five I
inspection packages also showed that the as-installed and as-inspected l conditions agreed, which indicated that both workmanship and inspection on the five hangers were adequate. Most of the rework required by the non-conformance reports (NCRs) found in the inspection packages consisted of filling in undersized fillet welds, although some rework was initiated by j design change authorization (DCA). The TRT learned that from July to September 1984, the NRC Region IV (RIV) inspectors reviewed a portion of the Unit 1 auxiliary feedwater system i while performing inspection 50-445/84-26. The RIV inspectors paid specific attention to two water lines which were connected to the condensate water i storage tank (CP-AFATCS-01) and were located in the safeguards tunnel. K-127
a They also inspected the pipe support in this area to the "as-built" vendor-certified drawing (VCD), and included critical dimensions of support members, weld size and type, support location, clearances, baseplates, workmanship, anchor bolt type and placement in their inspection. The RIV inspectors also reviewed the document package for the support. ' The RIV Inspectors inspected the following 10-inch supply line (AF-1-01-152-3) pipe supports / restraints identified on BRHL-AF-1-YD-002, and found no deficiencies or deviations. AF-1-001-020-Y33R Wall-mounted, single snubber AF-1-001-021-Y33K Floor-mounted, double snubber AF-1-001-025-Y33R Wall-mounted, rigid strut AF-1-001-028-Y43K Wall-mounted, double snubber AF-1-001-030-Y33R Ceiling-mounted, double strut i AF-1-001-036-Y33R Wall-mounted, sinqle strut i The RIV inspectors also examined the 8-inch return line (AF-1-035-152-3) to the condensate water storage tank and the following pipe supports / restraints, which are identified on BRHL-AF-1-YD-001, and found no deficiencies or deviations. 1 AF-1-035-001-Y33R Wall-or ceiling-mounted seismic pipe restraint , AF-1-035-003-Y33R Wall-or ceiling-mounted seismic pipe support AF-1-035-032-Y33R Wall-mounted seismic sway strut AF-1-035-034-Y33R Wall-mounted seismic sway strut AF-1-035-035-Y33K Wall-mounted double seismic snubber AF-1-035-037-Y33R Wall-mounted seismic sway strut The TRT determined that the RIV inspection identified no deficiencies such as those described by the alleger. In assessing this allegation, the TRT attempted to inspect the alleged deficient weld root gaps. The primary responsibility for a broader and more in-depth inspection of hangers belonged to other TRT Groups and, as described above, to the Region IV report. The TRT and a Brown & Root inspector went to the Unit 2 Safeguards Building tunnel and observed work in progress. Most of the piping had been installed, and approximately 12 pipe fitters were installing permanent hangers and snubbers on the tube steel. Fitups before welding appeared tight (less than 1/16-inch), straight, and uniform. The TRT observed no deficient fitups on the diagonal runs of the tube steel as described by the alleger. These findings are based on the technical assessment performed by the ' Miscellaneous Group; some of these findings appear to differ from those made by the TRT's QA/QC Group, which evaluated the same components from a different technical perspective.
- 5. Conclusion and Staff Positions: The TRT was unable to identify any improper fitups or gaps related to installed tube to baseplate weld-ments. The workmanship concerning the subject supports appeared to be .
good. Accordingly, this allegation has neither safety significance nor generic implications. K-128
3 a J ! The TRT was unable to provide the above findings and conclusions to the alleger because the alleger could not be located. Several attempts were ; made by letter and telephone to locate this alleger; however these attempts were unsuccessful. l-j 6. Actions Required: None. [ i I 1 I i I i i i l l 1 1 i d J DJ l i i i i l i 1 K-129
- 1. Allegation Category: Miscellaneous No. 14, Posting of NRC Form-3
- 2. Allegation Number: AM-19
- 3. Characterization: It is alleged that posting requirements for NRC Form-3 were not met during 1977-1982.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) reviewed the allegation, which was made in NRC Office of Investigation (01) document A4-83-005 dated May 20, 1984, and found no need to contact the alleger to further clarify the allegation.
The TRT reviewed the deposition given by Robert R. Taylor dated July 17, 1984, in which this NRC inspector stated that Texas Utilities Electric Company (TVEC) had posted a memorandum or letter in 1978 (about 6 months before Taylor became the resident inspector at the plant). This letter invited any site employee to contact the NRC if they had concerns about the quality of the construction of the plant. The NRC Region IV telephone number was the point of contact. This letter was the forerunner of the NRC Form-3 which became a posting requirement in October 1982. Mr. Taylor could not recall if the Form-3 was posted in October 1982, but had pre-pared a report in January 1983, which documented the posting. The Atomic Safety and Licensing Board (ASLB) deposition of C. Tedder, H. Hollis, C. Baker, and M. Hall, dated July 18, 1984, stated that NRC Form-3 had been posted from October 1982 until the present. The bulletin boards were periodically checked to assure proper posting. The TRT reviewed 10 CFR Part 50 dating back to 1976, and learned that the 10 CFR Part 50.7 requirement for posting the NRC Form-3 around sites under construction was not effective until October 12, 1982. The TRT observed the locations of the 12 Comanche Peak Steam Electric Station (CPSES) pro-ject bulletin boards currently in use, as well as 5 additional bulletin boards in various work spaces. The TRT determined that Form-3 was properly posted on all bulletin boards. The TRT interviewed the TUEC Radiation Protection Engineer (RPE), who currently is respondole for maintaining the CPSES Unit 1 bulletin boards, and the TUEC 6dministrative and control supervisor, who is now responsible for maintaining the 1,2 CPSES project bulletin boards. The RPE stated that TUEC designates official bulletin boards in work areas and other assembly areas and reviews them periodically to ensure compli- l ance with posting requirements. The TUEC administrative and control super-visor also stated that the CPSES project bulletin boards are reviewed perio-dically to ensure compliance with posting requirements and that locations for the bulletin boards may change as construction progresses. Management has not formally assigned responsibility in writing for establishing bulle-tin board locations or for maintenance and periodic review; the responsi- j bility is informally assumed. l The TRT also telephoned the Texas Utilities Service, Inc. (TUSI) personnel manager who was responsible for maintenance of the bulletin boards during the September 1982 to October 1983 period. The TUSI personnel manager stated that approximately five bulletin boards were in place and that an K-131
additional six bulletin boards were installed during his period of respon-sibility. This responsibility, however, was not a formally assigned job function.
- 5. Conclusion and Staff Positions: Based on a review of the NRC and TUEC !
depositions, interviews with the RPE, the TUEC administrative and control j supervisor and the TUSI personnel manager, and inspection of bulletin boards currently in place, the TRT concludes that letters were posted prior ; to October 1982, and that the NRC Form-3 was posted in a sufficient number of places to meet the intent of the applicable regulations after the posting l requirements became effective on October 12, 1982. Formal or written c 1 assignment of responsibility for NRC Form-3 posting could strengthen Since TUEC's i program if a policy of assigning responsibility were established. there was no requirement to post NRC Form-3 between 1977 and October, 1982, , and the form was posted for the balance of 1982 until the present, this 4 allegation is not substantiated. i I The TRT will provide the above findings and conclusions to the alleger by letter.
- 6. Actions Required: TUEC shall formally establish in writing the assignment l
of responsibility for posting and maintaining NRC Form-3 in prominent ; f locations. l
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i I 1 1 i i ) i l i i l i
- a !
i i K-132
- 1. Allegation Category: Miscellaneous 15, Drug Abuse
- 2. Allegation Number: AM-21
- 3. Characterization: In a letter dated March 7, 1984, it is alleged that there was widespread drug use and abuse at the Comanche Peak Steam Electric Station (CPSES) and that management did not give proper attention to the alleged problem.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) reviewed background material and an NRC Region IV report of inspection pursuant to temporary instruction (TI 2596/1) that documented discussions held with representatives of Texas Utilities Electric Company (TUEC) in early 1983. This report described TUEC's company policy on use or posses-sion of drugs and alcohol, employee assistance programs, background checks /
psychological tests, supervisory and employee awareness of drug / alcohol problems, and a drug / alcohol abuse detection program. During its review, the TRT learned that a drug abuse prevention program had been in effect at CPSES since 1974, when work on the project began, and that it included TUEC policy statements which emphasized that any employee possessing drugs or alcohol on company property was subject to immediate discharge. As a part of the review, the TRT interviewed managers, staff, and technicians affiliated with TUEC and Brown & Root (B&R). The staff also interviewed medical laboratory personnel and law enforcement officials in the CPSES area. The topics discussed in these interviews included the following: The methodology used in conducting drug investigations. The techniques used during pre-employment background investigations. The program for supervisors to ensure that they recognized employees with potential problems. The employee assistance program for permanent TUEC employees. The TRT determined through interviews and a review of personnel depart-ment practices and records that B&R corporate policy required prospective employees to take a physical examination to satisfy insurance or workmen's compensation requirements; however, the urinalysis given in this physical examination did not include an analysis for drugs. The TRT also learned that as a precondition of employment, both at TUEC and at Burns International (the CPSES physical security contractor), any employee requiring unescorted access to secured areas had to provide a urine specimen which was analyzed for a wide range of drugs. Prospective TUEC and security contractor employees were also subjected to an in-depth background investigation. TUEC's physical security plan commits them to mandatory screening and investigation of potential employees when the reactor becomes operational; however, TUEC elected to put these policies into effect prior to reactor operation. K-133
The TRT reviewed procedures and examined materials which revealed that, in anticipation of operation of CPSES, TUEC had initiated an indoctrination program for supervisors to aid them in recognizing unusual behavior caused by alcohol or drug abuse. Since B&R is the constructor and will not be involved with the operation of CPSES, there was no requirement in the TUEC Safeguards Security Plan concerning B&R personnel. Therefore, the B&R employees who were using drugs would not have been detected using this screening process. However, other screening measures, including the periodic use of dogs trained to detect drugs, were used to compensate for this lack of screening. The TRT interviewed the county sheriff and found that TUEC notified law enforcement authorities about their investigation regarding drug involve-ment by B&R employees and kept them advised of their findings. The measures described above were directed by TUEC and 8&R management, but despite these measures an incident occurred. The TRT learned that in early June 1984, TUEC security investigators from the corporate office in Dallas followed up on alleged onsite drug abuse by B&R employees, which reportedly occurred in the Unit 2 construction area and involved personnel from several trades and fields. TUEC began the investigation by interviewing ' the alleger and, using a networking approach, conducted a series of inter-views with 56 workers at CPSES. Following these interviews, TUEC security requested that 39 B&R employees take polygraph tests to support statements they had made when interviewed. This information was referred to the B&R personnel office and, as a result, 33 of the 39 employees who had been implicated by the interviews terminated employment. The TRT also interviewed the TUEC site QA manager and his special staff assistant concerning the use of drugs by quality assurance / quality control (QA/QC) personnel and its impact on safety-related work activities. They stated that in early June 1984, a TUEC investigator advised them that eight B&R QA/QC employees had been identified as being involved with drugs. They also stated that either three or four of the eight had left the project prior to the investigation, and the remainder terminated employment when the results of the investigation were referred to the B&R personnel office. Because inspections of safety-related work made by 8 of the 39 B&R employees involved with drugs may have been inadequate, a nonconformance report (M84-01840, dated June 15, 1984) was issued. This NCR addressed items / components in every system in Units 1 and 2, as those employees identified with drug involvement had worked in all areas of Unit 1 and 2. However, B&R interoffice memorandum (IOM), dated July 18, 1984, and TUEC IOM (TUQ-2289), dated August 14, 1984, provided justification te TUEC man-agement to exclude reinspection of the work of three B&R inspectors, either because an authorized nuclear inspector (ANI) had independently inspected this work and found no problems or because the work was not safety related. Therefore, the work of only five of the B&R inspectors involved in drugs was reinspected. The TRT found that in response to the NCR, TUEC QA personnel developed a reinspection program to assess the adequacy of those inspections which K-134
might have been inadequate. This program involved determining the total number of inspections for each of the inspectors and selecting a statisti-cal sampling plan from MIL-STD-1050, " Sampling Procedures and Tables for Inspection by Attributes." The sampling plan taken from MIL-STD-105D provided for a General Inspection, Level II, single, normal inspection with an acceptable quality level (AQL) of 4.0 considered to be adequate. The TRT reviewed the results of TUEC's reinspection program and found that the reinspected items / components were randomly selected and that a va'.id sample was reinspected by B&R inspectors. On August 31, 1984, the TUEC QA staff engineer stated that no significant deficiencies had been identified during their reinspection effort; however, eight minor deficiencies were referred to engineering in TUEC IOM QA-0047, dated September 21, 1984. IOM QA-0047 also included a request for an evaluation of the safety impli-cations of these minor deficiencies had they gone undetected. (The final sign off of this NCR was not completed as of January 31, 1985, pending VUEC engineering and legal review.) Following TUEC's reinspection program, the TRT randomly selected five per-cent of the TUEC sample to verify the adequacy of the reinspection. (The B&R inspectors involved in drug related activities were identified as A, B, C, 0, E, F, and G.) The TRT also included the work of the two inspec-tors whose work had not been reinspected by TUEC/B&R personnel. One B&R inspector's work was not included in this sample because it pertained to coatings, an area which was extensively inspected and evaluated by the TRT Coatings Group. Based on this sample, the TRT determined that the inspections were adequate. In addition, the TRT found no items / components that were deficient. The TUEC QA manager stated that although some craft personnel had been involved in drug abuse, their work was not reinspected because they were not responsible for final acceptance of their own work, but relied on in-process inspections and a final acceptance inspection made'by B&R inspec-tors who were not involved in the drug-related incident. In addition, an ANI inspected work done by craft personnel when ASME work was involved. . The TRT evaluated this position and reasoned that the sample of all inspec-tor's work would also include a sample of craft personnel work, and if no significant deficiencies were found, their justification woulj be accepted.
- 5. Conclusion and Staff Position: Based on the above review, the TRT con-cludes that TUEC had performed an investigation and identified B&R per-sonnel implicated by their refusal to take polygraph tests and their subsequent termination of employment. Although this allegation had potential safety significance and generic implications, TUEC wrote a nonconformance report which identified all work performed by the impli-cated B&R inspectors and reinspected by different inspectors. The rein-spection identified only minor deficiencies that have been referred to engineering for final evaluation and correction. This allegation appears to have some substance.
With respect to management, the TRT concluded that TUEC and site contrac-tor management and supervision had implemented strong measures to prevent K-135
drug use and abuse by CPSES personnel. In fact these commitments to Therefore, such a program exceed current NRC requirements and standards. there was no evidence that management did not give proper attention to the alleged problem to prevent drug use and abuse or deal with the incident that occurred. The TRT will provide its findings and conclusions to the appropriate group and to the alleger who was involved with this allegation.
- 6. Actions Required: TUEC shall provide a report of findings including the final engineering analysis of the minor deficiencies.
K-136
- 1. Allegation Category: Miscellaneous 16, Heating, Ventilating, and Air Conditioning System
- 2. Allegation Number: AM-22
- 3. Characterization: It is alleged that Texas Utilities Electric Company (TUEC) has not analyzed the heating, ventilating, and air conditioning system (HVAC) supports for seismic loads; that all HVAC components and supports inside the Containment Building were not properly considered in regard to their treatment as missiles; that the HVAC system is not properly supported; and, that HVAC failure during a postulated accident would allow the temperatures to rise to an unacceptable level inside the Containment Building.
- 4. Assessment of Safety Stanificance: The NRC Technical Review Team (TRT) found no need to contact the alleger to further clarify the allegation.
The TRT reviewed the Comanche Peak Steam Electric Station (CPSES) Final Safety Analysis Report (FSAR) to identify the HVAC's design and quality assurance requirements. FSAR Volume IV, Section 3.2, " Classification of Structures, Components and Systems," states that part of the containment ventilation system is seismic Category I; however, FSAR Volume XIV, Sec-tion 17.0, Appendix 17A, " List of Quality Assured Items," states that the c containment ventilation system (which contains eight subsystems / components) is seismic Category II and nonsafety related with the exception of the containment purge exhaust ductwork, supports, debris screen, and isolation valves, which are seismic Category I. Only the isolation valves, which are safety and code class 2, are safety related and seismic Category I. The TRT determined that the entire containment ventilation system is nonsafety related, except for the isolation valves referenced above. None of these nonsafety-related systems is necessary for the safe shutdown of the reactor or to prevent or mitigate the consequences of accidents or malfunctions in the reactor coolant pressure boundary. All components, except the containment purge exhaust, which are inside the Containment Building are seismic Category II, and need not operate during a safe shut-down earthquake (SSE) but simply are components which are not allowed to fall and damage an essential safety-related system. Therefore, the HVAC ductwork is not required to remove heat from containment. The system that removes heat from Containment Building is the containment spray system, which does not depend on HVAC duct work or HVAC supports. The allegation that temperatures would rise to an unacceptable level because of an inoperative HVAC is incorrect. The containment purge exhaust is classified as nonsafety related and seismic Category 1, which means it is designed to continue operating during a SSE; however, this ductwork system is not essential for the safe shutdown of the plant. The containment isolation valves close on a signal of high radiation to prevent a release to the environment, as specified by 10 CFR Part 20. This system is not used to remove heat from the Contain-ment Building either, as previously discussed. This is a containment spray system function so that the alleged high temperatures could not be caused ; by inoperative containment exhaust HVAC ductwork and supports. l K-137 l l i
- _ _ . _ _ . - _ _ _ _ _ _ . _ . _ _ _ _ _ _ - - _ _ _ _ _ - - _ _ _ - - _ _ - - _ _ _ _ _ - - _ _ _ _ _ _ _ =
The TRT reviewed FSAR Volume IV 5ection 3.5 and determined that TUEC had considered internally generated missiles inside the Containment Building. The allegation that the HVAC was not considered with respect to missiles is inqorrect. The TRT reviewed NRC Region IV (RIV) Inspection Report (IR) 50-445/83-24; 50-446/83-15 which documented a review of the allegations characterized above. It concluded that these allegations were without merit. This inspection based these conclusions on the review of the FSAR; an NRC Con-struction Appraisal Team report (CAT), dated April 11, 1983; and a special NRC inspection at Corporate Consulting & Development Company, LTD (CCL) the consultant responsible for HVAC design. The inspection of the con-sultant's analysis of design and seismic requirements, i.e., the seismic design techniques and assumptions, was acceptable. The TRT learned that RIV IR 50-445/84-16 documented a special inspection of the reactor Containment Building. Twenty-five duct supports segments in the Unit 1 containment air circulation and cooling system were inspected. The seismic supports were inspected to assure that installations were as designed or deviations were analyzed to assure the adequacy of the support of the HVAC systems. As a result of an interview with the alleger, additional inspection was performed and documented on December 18, 1984. The TRT randomly selected and observed various HVAC systems; the HVAC appeared to be properly sup-ported. The HVAC inside the Containment Building was analyzed and reported in CCL seismic analysis reports dated December 18, 1981, and July 24, 1984. The latter provided the following'Information: (1) the duct hangers were analyzed on a hanger-by-hanger basis, (2) the analysis was based on the latest as-built drawing, and (3) the hangers were designed and analyzed as frame structures having diagonal braces or without braces, thus relying on the bending of vertical supports to support lateral loads. The latter method may have caused the allegation that the HVAC was unsupported.
- 5. Conclusions and Staff Positions: Based on the review of design require-ments in the FSAR, the review of NRC inspections, and visual inspection of the HVAC systems, the TRT concludes that the allegations are based on the erroneous assumption that HVAC is required during a design basis accident.
The HVAC system has been properly designed and analyzed by an independent seismic consultant and by an analysis which included consideration of verti-cal and lateral supports needed to meet seismic Category 1 and 2 require-ments. Internally generated missiles inside containment were also analyzed. This allegation is not substantiated; therefore, it has neither safety significance nor generic implications. The TRT presented the above findings and conclusions to the alleger and agreed to recheck lateral supports at the alleger's request. After so doing, the TRT concluded that lateral supports were adequately considered.
- 6. Actions Required: None.
K-138
}
- 1. Allegation Category: Miscellaneous 17, Damage to Upper Internals
- 2. Allegation Number: AM-24
- 3. Characterization: It is alleged that damage occurred to the 15-foot by 2 1/2-inch stainless steel bars (subsequently determined by the NRC staff j to be thermocouple columns) located in the reactor vessel upper internal structures in the Unit 1 Reactor Building at Comanche Peak Steam Electric Station (CPSES). The alleger's concern is that approximately 1 foot from the top of the stainless steel bars, two of them were bent when they were struck by either a fork lift or a crane. The alleger contends that a rope pulled by a crane was then placed around the stainless steel bars and pulled in order to straighten them. It is further alleged that no docu-mentation was ever completed to show that this damage occurred.
- 4. Assessment of Safety Significance: NRC Region IV (RIV) inspected this allegation and documented the results in Inspection Report 50-445/84-08, 50-446/84-04 (July 26, 1984). Prior to this inspection (March 1984), the i RIV discussed this allegation with the alleger by telephone. There was also an NRC Office of Investigation (01) inquiry on this matter.
The NRC Technical Review Team (TRT) reviewed O! report QA-84-016 dated 1 April 11, 1984, the notes from the telephone conversation with the alleger, and NRC followup on Inspection Report 40-445/84-08, 50-446/84-04. In addi-tion, the TRT visually inspected the upper internal structures of the reac-tor vessel and reviewed a computerized index of CPSES documentation, i.e., nonconformance reports (NCRs) and/or procedures related to damage or prob-lems in the upper internals or reactor vessel head areas. Two documents, Westinghouse Field Deficiency Report (FOR) TBXM-10285 and Brown & Root (B&R) NCR M-11438, indicated that on October 14, 1983, the refueling cranc (a bridge crane that straddles the refueling cavity) was moved without crane interlocks or a " flagman," a condition which resulted in a bent thermocouple column. Although interlocks are normally used in such a case, at the time the alleged damage occurred no specific procedure was in effect that required use of interlocks. The TRT learned that when the alleged damage occurred, the upper core assembly was mounted on extension legs and was stored in its designated location in the refueling cavity. The extension legs elevated the upper internals so that the thermocouple column was in the refueling crane's normal path. Each of the four thermocouple-columns (tubes) is approxi-mately 17 feet long and provides support for the incore thermocouple tubing located between the upper core internals and the reactor vessel head; the bottom of each thermocouple column is attached to the upper core assembly. - The thermocouples in these columns are chromel-alumel wires that are threaded into guide tubes which penetrate the reactor vessel head through seal assemblies and terminate at the tcp end of the fuel assemblies. Thermocouple readings are monitored by a computer, with backup readout pro-vided by a precision indicator from the incore instrumentation, even if the
- computer is not in service. These thermocouples are not required for safety. (See Final Safety Analysis Report (FSAR], Section 7.7 " Control Systems Not Required for Safety.")
1 K-139 L __ __ -___ _ _
Westinghouse FDR TBXM-10285 and B&R NCR M-11438 indicated that thermocouple column R-11 (ID No.19546, Sub.1) on the reactor vessel internal structure was bent in an area approximately 2 feet above the support tube (Item 3 on Drawing 6116E84). The support tube had no apparent damage; however, the upper section of the thermocouple column and its respective protective sleeve were approximately 1 foot off the vertical. A review of records by the TRT indicated that, following the alleged damage, the thermocouple column had been properly aligned in a perpendicular direction. The area l was also visually inspected by the TRT for cracks with a ten power magnify-In addition, l ) ing glass. No cracking was observed where bending occurred. resistance readings were taken on February 6, 1984, with properly calibrated resistance instrumentation, and the results were acceptable. The TRT review of Deficiency Report TBXM 10285 and NCR M11438 indicated (1) the recommended corrective action of using a strong back and hydraulic rams to straighten the thermocouple action was reviewed and approved, (2) criteria (visual inspection, thermocouple resistance measurements) for acceptance of the corrective action were established, reviewed and approved, and (3) required QA/QC sign offs were completed. Between February and September 1984, Texas Utilities Electric Company (TVEC) personnel demonstrated proper operation of the manipulator crane, I its interlocks, and its safety features in accordance with Procedure . ICP-PT-40-03, " Manipulator Crane." Testing verified that the interlocks and safety features prevented any movement which would permit damage should procedures or personnel fail to perform as required during operation or j movement of the manipulator. Both the Westinghouse FDR and the B&R NCR indicated that TUEC identified and reported the damage to the steel bars, evaluated the potential and actual damage, and straightened the thermocouple column. Subsequent j evaluation by the TRT determined that the corrective action taken was 3 appropriate and adequate. In addition, the TRT inspected other areas, but identified no other documen-f! tation nor any physical evidence (dents, deep scratches, misalignment, or gouges) related to stainless steel or carbon steel with diameters between 1 and 4 inches on the reactor vessel head, the upper internals, or the core barrel.
- 5. Conclusion and Staff Positions: The TRT determined that the stainless steel bars (thermocouple columns) were bent by the refueling crane and corrected by the recommended action of using a strong back and hydraulic ram (s). Further review indicated that the thermocouples are not safety related, however. The TRT concludes that TUEC's reporting and corrective actions were appropriate for this type of equipment damage, and that an The appropriate level of quality was applied to the corrective actions.
corrective actions taken indicate that no additional repairs or potential deleterious equipment failure should result from the bending of the thermo-couple column and subsequent straightening. Accordingly, this allegation has neither safety significance nor generic implications. l K-140
On November 1, 1984, the TRT provided the above findings and conclusions to the' alleger. The alleger stated that his questions or concerns were answered and he had no further concerns.
- 6. Actions Required: None.
l' t 8 i K-141
- 1. Allegation Category: Mi;cellaneous 18, Broken Internal Wires in Polar Crane Festooned Cable and Crane Movement Interference
- 2. Allegation Number: AM-25
- 3. Characterization: It is alleged that internal wires were broken in the polar crane festooned cables and that the polar crane hit unspecified hangers while operating.
- 4. Assessment of Safety Significance: The alleger was interviewed by the NRC Technical Review Team (TRT) on August 3, 1984, to obtain additional infor-mation regarding the allegation. On August 30, 1984, the TRT and a Texas Utilities Electric Company (TVEC) quality control (QC) inspector visually examined the festooned cables. There was no visible damage on any of the cables. In addition, a review of preoperational inspection megger test data sheets revealed that all tests were satisfactory.
The TRT visually inspected the polar crane during three rotations to deter-mine if there were any interferences between the crane and supports or other installed items, and noted no interferences. It is possible that the alleger was referring to the problem of the uplift lugs striking the crane girder stiffener plates, which is described in nonconformance report (NCR) M-81-00064; however, this problem was resolved in accordance with DCA 11311, Rev. 1. On September 12, 1984, the TRT, accomp'anied by a Brown & Root (B&R) QC electrical inspector and electrician, opened the two electrical junction boxes that feed the festoons on the polar crane walkway. The inspector visually inspected all of the wires in both boxes and found no broken or non-terminated wires. The TRT asked the operator of the crane about pro-blems with the crane, specifically asking if the limit switches cut out properly. The crane operator then demonstrated the operation of the bridge crane, running it until the limit switch cut out and a signal light indi-cated that it cut out. Again, he stated that there were no problems with the crane. The TRT determined from a B&R QC inspection that no records or nonconform-ance reports existed which may have documented the alleged defective festooned cables because they were classified as nonsafety related. The TRT found only the records for megger testing which were previously discussed.
- 5. Conclusion and Staff Positions: The TRT found no damaged festooned cables.
However, the polar crane uplift lugs did strike the crane girder stiffener plates and this had potential safety significance and generic implications. The TRT did find an NCR documenting the damaged plates corrective action, and the TRT verified that corrective action was taken, i.e., the crane now operates without such interference. Based on a review of applicable docu-mentation, examination of the polar crane cables and wiring on the polar crane walkway, and interviews with the crane operator, the TRT concludes that this allegation was substantiated; however, appropriate corrective action was taken. Accordingly, this allegation has neither safety signifi-cance nor generic implications. K-143
i i On November 1, 1984, the TRT provided the above findings and conclusions to the alleger. The alleger was satisfied with the findings and had no further concern regarding this matter. However, during this interview the i alleger brought up a commercial concern regarding premature replacement of
- the cable. Although this is not related to a safety issue, the NRC's j'
Senior Resident Inspector stated he would review the matter during a i future routine inspection.
- 6. Actions Required: None.
i 4 4 .I 4 i i i i i 1 i
' K-144
i l I
- 1. Allegation Category: Miscellaneous 19, Chloride Contamination of Radwaste System Piping i 2. Allegation Number: AM-30 V
- 3. Characterization: It is alleged that workers habitually urinated on stainless steel pipe located in the radwaste system.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) found no need to contact the alleger to further clarify the allegation.
The TRT reviewed background information pertinent to this allegation and searched the Texas Utilities Electric Company's (TUEC) quality assurance records relating to piping cleanliness. Nonconformance report (NCR) M-82-00305 described an instance where piping in the radwaste area (above the waste monitor tanks in Room 2) was " contaminated by unknown liquid substances of unspecified chemical composition." The NCR stated that hold tags were applied and lines 2 WP-X-218-151-R5 and 2 WP-X-208-151-RS, located in the Unit 1 Auxiliary Building (elevation 790 feet of Gibbs & Hill drawing 2323-Al-0507, Revision 9), were subsequently cleaned and swipe-tested according to Procedure QI-QP-11.1-65 to assure that the sur-faces were free of chlorides and fluorides. TUEC verified corrective action and closed this NCR on May 27, 1982. The TRT discussed this NCR with TUEC quality and engineering personnel, who stated that the NCR quoted a report made by a QC inspector who wit-nessed a worker urinating on the piping. TUEC personnel further stated that they knew of no other similar instances; however, all safety-related stainless steel piping surfaces (outside) are routinely cleaned prior to final turnover. - The safety significance related to chlorides, a chemical present in human urine, on stainless steel surfaces depends on the service conditions and residual stresses that may be present. If excessive stress (near the yield strength) and chloride contamination are present, the alloy may fail because of stress corrosion cracking. Since the alloy is expected to operate with a design load applied, it is necessary to ensure that chlorides are not present. In this case, the piping was cleaned to remove any chlorides that could have been deposited by urine. Thermal insulation is applied after cleaning and this protects safety-related piping (necessary for safe shut-down) from further contamination. The radwaste piping above the radwaste tanks is nonsafety related and is not needed for the safe shutdown of the plant. On August 29, 1984, the TRT inspected the radwaste areas (Rooms 179, 184, and 185) and found them locked and access to them controlled. Housekeeping appeared to be excellent, and the TRT detected no odors which might indicate that the area was further contaminated. The number of craft personnel who work in the Unit 1 buildings has been limited for.several months, as com-pared to earlier periods, because this unit is virtually completed. Because limited work is in progress and personnel access controls are in place, it appears unlikely that other similar incidents occurred after cleaning. I K-145
The TRT randomly selected safety-related piping (lines RC-1-052-2501-R-1, 27.5 ID and 3/4-MS-1-194-1501-2) and reviewed the records to determine if the piping had been cleaned and swipe-tested in accordance with Gibbs & Hill Inc. Specification 2323-MS-100, Revision 8, and Brown & Root Inc. Procedures CP-QP-11.12, Revision 16 and QI-QP-11.1-65, Revision 4. Surface contamination reports J479, J492, and J497 document test results that show both the chloride and fluoride content are below the maximum specified limit of 0.0015 mg/dm. In addition, Region IV inspectors observed the . external cleanliness of the reactor coolant system piping as part of their May 14 through June 20, 1984, inspection (documented in RIV Inspection Report 50-445/84-16) and identified no deviations or violations of requirements. The TRT found no evidence to support that this incident occurred in any other area. During construction, toilet facilities are not always close to each work area; therefore, workers do sometimes urinate in unauthorized areas. However, the evidence indicates that all safety-related piping is cleaned and tested before being placed into service, eliminating potential contamination.
- 5. Conclusion and Staff Positions: The TRT found that an NCR was written on radwaste piping because a QC inspector saw a worker urinating on this piping. This allegation had potential safety significance and generic implications because the incident may have involved safety-related piping.
The radwaste piping which was contaminated was subsequently cleaned. The TRT found no other instances where this happened; however, TUEC's proce-dures for maintaining chloride and fluoride surface contamination levels below specified limits appear to be acceptable, were followed, and will eliminate the contamination of critical safety-related piping, whether the incident was isolated or habitual. Moreover, the radwaste piping is nonsafety related. Accordingly, this allegation has neither safety significance nor generic implications. The TRT will provide written feedback to the alleger describing its find-ings and conclusions.
- 6. Actions Required: None.
d K-146 i
1
- 1. Allegation Category: Miscellaneous 20, No Procedures or Guidance Provided for Rigging and Handling large Components / Equipment
- 2. Allegation Number: AM-23(a)
- 3. Characterization: An NRC Region IV Resident Inspector identified a vio-lation as a result of a discussion with a craft person who stated that he had not received instructions about how to rig and handle a large motor-operated valve.
- 4. Assessment of Safety Significance: The NRC Technical Review Team (TRT) found no need to contact the alleger to further clarify the allegation.
The TRT reviewed NRC Inspection Report 50-445/79-27, 50-446/79-26 and its corresponding Notice of Violation (NOV). The TRT also reviewed the Texas Utilities Electric Company (TUEC) response to these documents (TXX-3080, dated December 18, 1979), which stated that the subject valve was not mis-i handled, nor was it damaged. The engineering organization had not, how-
; ever, reviewed specific vendor rigging or handling recommendations or noted the procedures for loads exceeding 2000 pounds. An NRC followup inspec-
] tion verified that Brown & Root (B&R) Procedures CP-CPM-6.3, 35-1195-CCP-24, 35-1195-ACP-3, and QI-QAP-13.1-1 were reviewed by TUEC and revised appro-priately. NRC Inspection Report 50-445/80-18, 50-446/80-18 (dated September 19, 1980) documented corrective action during the followup in_spection. The TRT interviewed TUEC's Rigging Craft Superintendent, Assistant Mechan-ical Superintendent, and Senior Staff Engineer. They stated that the revised procedures (specifically, CCP-2A, Revision 4, " Rigging"; CP-CPM-6.3, Revision 10, " Preparation, Approv.1, and Control of Operation Travelers"; and, CP-CPM-6.9, Revision 2, " General Piping Procedure") adequately con- . trolled heavy lifts of equipment and components. Nonconformance report
- (NCR) M-2128 documented the problem which was identified as a viola-tion, and the appropriate site personnel reviewed the NRC inspection report and concurred with the corrective action. In addition, the TRT independ-ently reviewed tte revised procedures for the control of heavy lifts of equipment and found the control of rigging and handling to be acceptable for loads less than or exceeding 2000 pounds.
- 5. Conclusion and Staff Positions: The TRT determined that Region IV (RIV) confirmed that the craf tperson's stated need for better instructions was correct and confirmed followup inspection by the RIV inspector to verify that corrective action was accomplished in accordance with TUEC letter TXX-3080 (December 18, 1979). The TRT concludes that the failure to-pro-vide proper instructions for rigging and handling heavy loads is safety significant and has generic implications; however, corrective action was taken. No evidence of further inadequacies in this area was found; con-sequently the allegation requires no further action.
l The TRT tried to provide the above findings and conclusions to the
- alleger; however, the alleger's identification is unknown.
I 4
- 6. Actions Required: None.
i J K-147
Attachment 3 49 b UNITED STATES / ., \ NUCLEAR REGULATORY COMMISSION wasmusToN. O C. 20555 s, , iSi,? 18 ON Dockets: 50-445 50-446 Texas Utilities Electric Compar.y Attn: M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
SUBJECT:
COMANCHE PEAK REVIEW On July 9, 1984, the staff began an intensive onsite effort designed to complete a portion of the reviews necessary for the staff to reach its decision regarding the licensing of Comanche Peak Unit 1. The onsite effort covered a number of areas, including allegations of improper construction practices at the facility. The NRC assembled a Technical Review Team (TRT) responsible for evaluating most of the technical issues at Comanche Peak, including allegations. The TRT has recently identified a number of items that have potential safety implications for which we require additional information. These items are listed in the enclosure to this letter. Further background information regarding these' issues will be published in a Supplement to a Safety Evaluation Report (SSER), which will document the overall TRT's assessment of the significance of the issues examined. The items in the enclosure to this letter, which are in the general areas of electrical / instrumentation, civil / structural and test programs, cover only a portion of the TRT's effort. The TRT evaluation of items in the areas of mechanical, QA/QC, and coatings, and its consideration of the programmatic implications of these findings, are still is progress. A summary of these issues will be provided to you at a later date. You are requested to submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified. This program plan and its implemen-tation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic implic-ations on safety-related systems, programs, or areas. The collective significance of these deficiencies should also be addressed. Your program plan should also include the proposed TUGC0 action to assure that such problems will be precluded from occurring in the future. K-149
l l SEP 18 aa4 Mr. M. D. Spence This reouest is submitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding information/ evaluation needs that could potentially affect the safe operation of their plant. Further requests for additional information of this nature will be made, if necessary, as the activities of the TRT progress. Sincerely, t .- j',
'eh G. 'Eisenhut,'DiNetor Division of Licensing, NRR
Enclosure:
As stated cc w/ enclosure See next page R K-150
COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Company . 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels & Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B . 81 Dallas, Texas 75201 B. R. Clements Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B.'81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq. P. O. Box 355 1200 New Hampshire Avenue, N. W. Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountabi.lity Project Austin, Texas 78711 1901 Que Street, N. W. Washington, D. C. 20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq. Energy Orrick, Herrington & Sutcliffe ; 1426 South Polk 600 Montgomery Street ! Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq. CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W. San Francisco, California 94111 Suite 611 Washington, D. C. 20036 K-151 i .
~
ENCLOSURE 1 REQUEST FOR ADDITIONAL INFORMATION I. Electrical / Instrumentation Area
- a. Electrical Cable Terminations The Technical Review Team (TRT) inspected random samples of safety-related terminations, butt splices inside panels, and vendor-installed terminal lugs in General Electric (GE) motor control centers, and reviewed documentation relative to the installations.
- 1. The TRT found a lack of awareness on the part of quality control (QC) electrical inspectors to document in the inspection reports when the installation of the " nuclear heat-shrinkable cable insulation sleeves" was required to be witnessed.
Accordingly, TUEC shall clarify procedural requirements and provide additional inspector training with respect to the areas in which nuclear heat-shrinkable sleeves are required on splices and assure that such sleeves.are installed where required.
- 2. The TRT found inspection reports that did not indicate that the required witnessing of splice installation was done. Examples are as follows:
IR ET-1-0005393 IR ET-1-0005396 IR ET-1-0005394 IR ET-1-0006776 IR ET-1-0005395 IR ET-1-0014790 Accordingly, TUEC will assure that all QC inspections requiring witnessing for butt splices have been performed and properly documented; and verify that all butt splices are properly identified on the appropriate drawings and are physically identified within the appropriate panels.
- 3. The TRT found a lack of splice qualification requirements and provisions in the installation procedures to verify the operability of those circuits for which splices were being used.
Accordingly, TUEC shall develop adequate installation / inspection procedures to assure that the wiring splicing materials are qualified for the appropriate service conditions, and that splices are not located adjacent to each other.
- 4. Selected cable terminations were found that did not agree with their locations on drawings. Examples are as follows:
K-152
Panel CP1-ECPRCB-14, Cable E0139880 Panel CPI-ECPRTC-16, Cable E0110040 Panel CP1-ECPRTC-16, Cable E0118262 Panel CP1-ECPRTC-27, Cable EG104796 Panel CPX-ECPRCV-01, Cable EG021856 Panel CP1-ECPRCB-02, Cable NK139853 (nonsafety) Accordingly, TUEC shall reinspect all safety-related and associated terminations in the control room panels and in the termination cabinets in the cable spreading room to verify that their locations are accurately depicted on drawings. Should the results of this reinspection reveal an unacceptable level of nonconformance to drawings, the scope of this reinspection effort shall be expanded to include all safety-related and associated terminations at CPSES.
- 5. The TRT found cases where nonconformance reports (NCRs) concerning vendor-installed terminal lugs in GE motor control centers had been improperly closed. Examples are NCR Nos.
E-84-01066 through NCR E-84-01076, inclusive. Accordingly, TUEC shall reevaluate and redisposition all NCRs related to vendor-installed terminal lugs in GE motor control centers.
- b. Electrical Equipment Separation The TRT reviewed the separation criteria between separate cables, trays and conduits in the main control room and cable spreading room in Unit 1, and the compatibility of the electrical erection specifications with regulatory requirements. The TRT reviewed documentation and inspected random samples of separation between safety-related cables, trays and conduits and between them and nonsafety-related cables, trays and conduits.
- 1. In numerous cases, safety-related cables within flexible conduits inside main control room panels did not meet minimum separation requirements. Examples are as follows:
Panel CP1-EC-PRCB-02 Panel CP1-EC-PRCB-07 Panel CP1-EC-PRCP-06 Panel CP1-EC-PRCB-08 Panel CP1-EC-PRCB-09 Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room for Unit 1, that contain redundant safety-related cables within conduits, or safety and non-safety related cables within conduits, and either correct each violation of the separation criteria, or l K-153 , i l 4 l
demonstrate by analysis the acceptability of the conduit as a barrier for each case where the minimum separation is not met.
- 2. In several cases, separate safety and nonsafety-related cables and safety and nonsafety-related cables within flexible conduits inside main control room panels did not meet minimum separation requirements (Table 1 identifies examples of these cases). No evidence was found that justified the lack of separation.
Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room of Unit 1, and either correct each violation of the separation criteria concerning separate cables and cables within flexible conduits, or demonstrate by analysis the adequacy of the flexible conduit as a barrier.
- 3. The TRT found that the existing TUEC analysis substantiating.the adequacy of the criteria for separation between corduits and cable trays had not been reviewed by the NRC staff.
Accordingly, TUEC shall submit the analysis that substantiates the acceptability of the criteria stated in the electrical erection specifications governing the separation between independent conduits and cable trays.
- 4. The TRT found two minor violations of the separation criteria inside panels CP1-EC-PRCB-09 and CPI-EC-PRCB-03 concerning a barrier that had been removed and redundant field wiring not meeting minimum separation. The devices involved with the barrier were FI-2456A, PI-2453A, PI-2475A, and IT2450, associated with Train A; and FI-2457A, PI-2454A, PI-2476A, and IT-2451, associated with Train B. The field wiring was associated with devices HS-5423 of Train B and HS-5574, nonsafety-related.
Accordingly, TUEC shall correct two minor violations of the ! separation criteria inside panels CPI-EC-PRCB-09 and l CP1-EC-PRCP-03 concerning a barrier that had been removed and : redundant field wiring not meeting minimum separation. ,
- \
K-154,
Table 1 Examples of Cases of Safety or Nonsafety-Related Cables In Contact With Other Safety-Related Cables Within Conduits in Control Room Panels 1.' Control Panel CPI-EC-PRCB Containment Spray System Cable No. Train Related Instrument EG139373 8( IJndetermined E0139010 A (green) orange) Undetermined
- 2. Control Panel CP1-EC-PRCB Reactor Control System Cable No. Train Related Instrument EG139383 ET9reen) Reactor manual trip switch E0139311 A (orange) Undetennined
- 3. Control Panel CP1-EC-PRCP Chemical & Volume Control System Cable No. Train Related Instrument EG139335 Egreen) LCV-112C E0139301 A (orange) Undetennined
- 4. Control Panel CP1-EC-PRCB Auxiliary Feedwater Control System Cable No. Train Related Instrument E0139753 A (orange) FK-2453A E0139754 A (orange) FK-2453B E0139756 B(green) FK-2454A EG139288 8 (green) FK-2454B I
1 K-155
j
- c. Electrical Conduit Supports The TRT examined the nonsafety-related conduit support installation in selected seismic Category I areas of the plant. The support installation for non-safety related conduits less than or equal to 2 inches was inconsistent with seismic requirements and no 4
evidence could be found that substantiated the adequacy of the installation for nonsafety-related conduit of any size. According to Regulatory Guide 1.29 and FSAR Section 3.78.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the function of safety-related components or cause injury to plant personnel. 2 Accordingly, TUEC shall propose a program that assures the adequacy of the seismic support system installation for nonsafety-related
, conduit in all seismic Category I areas of the plant as follows:
- 1. Provide the resul.ts of seismic analysis which demonstrate that
; all nonsafety-related conduits and their support systems, .
satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. Verify that nonsafety-related conduits less than or equal to 2 inches in diameter, not installed in accordance with the requirements of Regulatory Guide 1.29, satisfy applicable design requirements.
- d. Electrical QC Inspector Training / Qualifications The TRT examined electrical QC inspector training and certification files, and requirements for personnel testing, on-the-job training, ,
and recertification. The TRT also interviewed selected electrical QA/QC personnel. l
- 1. The TRT found a lack of supportive documentation regarding
' personnel qualifications in the training and certification files, as required by procedures and regulatory requirements.
Also, the TRT found a lack of documentation for assuring that the requirements for electrical QC inspector recertification were being met. Specific examples are:
- One case of no documentation of a high school 4
diploma or General Equivalency Diploma. J f ' K-156
- One case of no documentation to waive the remaining 2 months of the required 1 year experience.
- One case where a QC technician had not passed the required color vision examination administered by a professional eye specialist. A makeup test using colored pencils was administered by a QC supervisor, was passed, and then a waiver was given.
- Two cases where the experience requirements to become a Level 1 technician were only marginally met.
- One case of no documentation in the training and certification files substantiating that the person met the experience requirements.
Accordingly TUEC shall review all the electrical QC inspector training, qualification, certification and recertification files against the project requirements and provide the information in such a form that each requirement is clearly shown to have been met by each inspector. If an inspector is found to not meet the training, qualification, certification, or recertification requirements, TUEC shall then review the records to determine the adequacy of inspections made by the unqualified individuals and provide a statement on the impact of the deficiencies noted on the safety of the project.
- 2. The TRT found a lack of guidelines and procedural requirements for the testing and certifying of electrical QC inspectors. Specifically, it was found that:
No time limit or additional training requirements existed between a failed test and retest. No controls existed to assure that the same test would not be given if an individual previously failed that test. No consistency existed in test scoring. No guidelines or procedures were available to control the disqualifj ation of questions from the test. No program was available for establishing new tests (except when procedures changed). The same tests had been utilized for the last 2 years. Accordingly, TUEC shall develop a testing program for electrical QC inspectors which provides adequate administrative guidelines, procedural requirements and test flexibility to assure that suitable proficiency is achieved and maintained. l K-157
The deficiencies identified with the electrical QC inspections have generic implications to other construction disciplines. The implications of these findings will be further assessed as part of the overall programmatic review of QC inspector training and qualification and the results of this review will be reported under the QA/QC category on " Training and Qualification."
,I I . Civil / Structural Area
- a. Unable to Justify Reinforcing Steel Omitted in the Reactor Cavity The TRT investigated a documented occurrence in which reinforcing steel was omitted from a Unit I reactor cavity concrete placement between the 812-foot and 819-foot 1-inch elevations. This reinforcement was installed and inspected according to drawing 2323-51-0572, Revision 2. However, after the concrete was placed, Revision 3 to the drawing was issued showing a substantial increase in reinforcing steel over that which was installed. Gibbs & Hill Engineering was informed of the omission by Brown & Root Nonconformance Report CP-77-6. Gibbs & Hill Er.gineering replied that the omission in no way inpaired the structural integrity of the structure. Nevertheless, the' additional reinforcing steel was added as a precaution against cracking which might occur in the vicinity of the neutron detector slots should a loss of coolant accident (LOCA) occur. A portion of the omitted reinforcing steel was also placed in the next concrete lift above the 819-foot 1-inch level. This was done to partially compensate for the reinforcing steel omitted in the previous concrete lift and to minimize the overall area potentially subject to cracking.
The TRT requested documentation indicating that an analysis was performed supporting the Gibbs & Hill conclusion. The TRT was subsequently informed that an analysis had not been performed. Therefore, the TRT cannot determine the safety significance of this issue until an analysis is performed verifying the adequacy of the i reinforcing steel as installed, r Accordingly, TUEC shall provide an analysis of the as-built condition of the Unit I reactor cavity that verifies the adequacy of the ; reinforcing steel between the 812-foot and 819-foot i-inch elevations. The analysis shall consider all required load combinations. b ., Falsification of Concrete Compression Strength Test Results The TRT investigated allegations that concrete strength tests were falsified. The TRT reviewed an NRC Re Report No. 50-445/79-09; 50-446/79-09)gion of this matter IV that investigation included (IE , i K-158.
l interviews with fifteen individuals. Of these, only the alleger and one other individual stated they thought that falsification occurred, but they did not know when or by whom. The TRT also reviewed slump and air entrainment test results of concrete placed February during)the 1977 andperiod did notthe findalleger was employed any apparent variation (January in the 1976 to uniformity of the parameters for concrete placed during this period. Although the uniformity of the concrete placed appears to ninimize the likelihood that low concrete strengths were obtained, other allegations were raised concerning the falsification of records associated with slump and air content tests. The Region IV staff addressed these allegations by assuming that concrete strength test results were adequate. Furthermore, a number cf other allegations dealing with concrate placement problems (such at deficient aggregate grading and concrete in the mixer too long) were also resolved by assuming that concrete strength test results were adequate. The TRT agrees with. Region IV that, while the preponderance of evidence i suggests that falsification of results did not take place,
! the matter cannot be resolved completely on the basis of concrete
- strength test results, especially if there is any doubt about whether j they may have been falsified. Due to the importance of the concrete l strength test results, the TRT believes that additional action by TUEC is necessary to provide confimatory evidence that the reported
} concrete strength test results are indeed representative of the strength of the concrete installed in the Category I concrete
, structures.
Accordingly, TUEC shall determine areas where safety-related concrete was placed between January 1976 and February 1977, and provide a j program to assure acceptable concrete strength. The program shall include tests such as the use of random Schmidt hammer tests on the concrete in areas where safety is critical. The program shall include a comparison of the results with the results of tests per-formed on concrete of the same design strength in areas where the strength of the concrete is not questioned, to determine if any significant variance in strength occurs. TUEC shall submit the v/ogram for performing these tests to the NRC for review and approval prior to performing the tests.
- c. Maintenance of Air Gap Between Concrete Structures 4
The TRT investigated the requirements to maintain an air gap between i concrete structures. Based on the review of available inspection
- reports and related documents, on field observations, and on discussions with TUEC engineers, the TRT cannot detemine whether an adequate air gap has been provided between concrete structures. Field investigations by B&R QC inspectors indicated unsatisfactory conditions due to the presence of debri,s in the air K-159
l gap, such as wood wedges, rocks, clumps of concrete and rotofoam. The disposition of the NCR relating to this matter states that the
" field investigation reveals that most of the material has been removed." However, the TRT cannot detemine from this report (NCR C-83-01067) the extent and location of the debris remaining between the structures.
Based on discussions with TUEC engineers, it is the TRT's understanding that field investigations were made but that no pemanent records were maintained. In addition, it is not apparent that the permanent installation of elastic joint filler material ("rotofoam") between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the seismic analysis assumptions and dynamic models used to analyze the buildings, as these analyses are delineated in the Final Safety Analysis Report (FSAR). The TRT, therefore, concludes that TUEC has not adequately demonstrated compliance with FSAR Sections 3.4.1.1.1, 3.8.4.5.1, and 3.7.B.2.8, which require i separation of Seismic Category I buildings to prevent seismic interaction during an earthquake. Accordingly, TUEC shall:
- 1. Perfom an inspection of the as-built condition to confirm that adequate separation for all seismic category I structures has been provided.
- 2. Provide the results of analyses which demonstrate that the presence of rotofoam and other debris between all concrete structures (as detemined by inspections of the as-built conditions) does not result in any significant increase in seismic response or alter the dynamic response characteristics of the Category I structures, components and piping when compared with the results of the original analyses,
- d. Seismic Design of Control Room Ceiling Elements The TRT investigated the seismic design of the ceiling elements installed in the control room. The following matrix designates those ceiling elements present in the control room and their seismic category designation:
K-160 l l
- 1. Heating, Ventilating and Air Conditioning - Seismic Category I
- 2. Safety-Related Conduits - Seismic Category I
- 3. Nonsafety-Related Conduits - Seismic Category II
- 4. Lighting Fixtures - Seismic Category II
- 5. Sloping Suspended Drywall Ceiling - Non-Seismic
- 6. Acoustical Suspended Ceiling - Non-Seismic
- 7. Lowered Suspended Ceiling - Non-Seismic According to Regulatory Guide 1.29 and FSAR Section 3.78.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the functions of safety-related components or cause injury to operators.
For the nonseismic items (other than the sloping suspended drywall ceiling), and for nonsafety-related conduits whose diameter is 2 inches or less, the TRT could find no evidence that the possible effects of a failure of these items had been considered. In addition, the TRT determined that calculations for seismic Category II components (e.g., lighting fixtures) and the calculations for the sloping suspended drywall ceiling did not adequately reflect the rotational interaction with the nonseismic items, nor were the fundamental frequencies of the supported masses detennined to assess the influence of the seismic response spectrum at the control room ceiling elevation would have on the seismic response of the ceiling elements. Accordingly, TUEC shall provide:
- 1. The results of seismic analysis which demonstrate that the nonseismic items in the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
- 2. An evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems.
- 3. Verification that those items in the control room ceiling not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
- 4. The results of an analysis that justify the adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
K-161 l _ - . ._. . _ _
l l
- 5. The results of an analysis which demonstrate that the foregoing problems are not applicable to other Category II and nonseismic structures, systems and components elsewhere in the plant.
- e. Unauthorized Cutting of Rebar in the Fuel Handling Building The TRT investigated an alleged instance of unauthorized cutting of rebar associated with the installation of the trolley process aisle rails in the Fuel Handling Building. The claim is that during installation of 22 metal plates in January 1983, a core drill was used to drill about 10 holes approximately 9 inches deep. The TRT reviewed the reinforcement drawings for the Fuel Handling Building and determined that there were three layers of reinforcing steel in the top reinforcement layer of the slab. This reinforcement layer consisted of a No. 18 bar running in the east-west direction in the first and third layers, and a No. 11 bar running in tne north-south direction on the second layer. The review also revealed that the layout of the reinforcement and the trolley rails was such that the east-west reinforcement would interfere with the drilling of holes along only one rail location. However, if 9-inch holes were drilled, both the first and third layers of No. 18 reinforcement would be cut.
Design Change Authorization No. 7041 was written for authorization to cut the uppermost No.18 bar at only one rail location, but did not reference authorization to cut the lower No. 18 bar. DCA-7041 also stated that the expansion bolts and base plates may be moved in the east-west direction to avoid interference with reinforcement running in the north-south direction. The infonnation, described in DCA-7041, was substantiated by Gibbs & Hill calculations. If the ten holes were actually drilled 9 inches deep, then the allegation that the reinforcement was cut without proper authorization would be valid. Accordingly, TUEC shall provide:
- 1. Information to demonstrate that only the No.18 reinforcing steel in the first layer was cut, or
- 2. Design calculations to demonstrate that structural integrity is maintained if the No. 18 reinforcing steel on both the first and third layers was cut.
III. Test Programs Area
- a. Hot Functional Testing (HFT)
The TRT reviewed a sample of the completed data packages for HFT preoperational test procedures, pertinent startup ar.ninistrative procedures, NRC inspection reports, and the preoperational test index and its schedule. The TRT also inspected test deficiency reports K-162
(TDRs) that were generated as a result of test deficiencies found prior to and during HFT.
- 1. Chapter 14 of the FSAR and Regulatory Guide 1.68 provide requirements for the conduct of preoperational testing.
In reviewing test data packages, the TRT found that certain test objectives were not met. It appears that the Joint Test Group approved incomplete data packages for at least three preoperational hot functinal tests. These were: Test Procedure Deficiency ICP-PT-02-12, " Bus Because acceptable voltages Voltage and Load Survey" could not be achieved with the specified transformer taps, they were changed. A subsequent engineering evaluation required returning to the original taps, but no retest was performed. ICP-PT-34-05, " Steam Level detectors 1-LT-517, 518 Generator Narrow Range and 529 were rep, laced with Level Verification" temporary equipment of a design that was different from that which was to be eventually installed I 1CP-PT-55-05 Level detector 1-LT-461 appeared
" Pressurizer Level to be out of calibration during the Control" test and was replaced after the test.
The retest approved by the JTG was a cold calibration rather than a test consistent with the original test objective, which was to obtain satisfactory data under hot conditions. Accordingly, TUEC shall review all complete preoperational test data packages to ensure there are no other instances where test objectives were not met, or prerequisite conditions were not satisfied. The three items identified by the TRT shall be included, along with appropriate justification, in the test deferral packages presented to the NRC. K-163
- 2. The TRT noted during a review of HFT completed test data that the JTG did not approve the data until after cooldown from the test. The tests are not considered complete until this approval is obtained. In order to complete the proposed post-fueling, '
deferred preoperational HFT, the JTG, or a similarly qualified group, must approve the data prior to proceeding to initial I criticality. The TRT did not find any document providing > assurance that TUEC is committed to do this. Accordingly, TUEC shall consnit to having a JTG, or similarly qualified group, review and approve all post-fueling preoperational test results prior to declaring the system operable in accordance with the technical specifications.
- 3. The TRT pointed out that in order to conduct preoperational
.l tests'at the necessary temperatures and pressures after fuel load, certain limiting conditions of the proposed technical specifications cannot be met, e.g., all snubbers will not be operable since some will not have been tested. Accordingly, TUEC shall evaluate the required plant conditions ; i for the deferred preoperational tests against limiting conditions in the propos(d technical specifications and obtain ! NRC approval where deviations from the technical specifications t are necessary. l
- 4. Data for the thermal expansion tests (which have not yet been l approved by the JTG) did not provide for traceability between !
the calibration of the measuring instruments and the monitored locations, as required by Startup Administrative Procedure-7. - I The information was separately available in a personal log held . j by Engineering. ; Accordingly, TUEC shall incorporate the information necessary to l provide traceability between thermal expansion test monitoring locations and measuring instruments. TUEC shall also establish : administrative controls to assure appropriate test and measuring l equipment traceability during future testing. { j b. Containment Intergrated Leak Rate Testing (CILRT) The TRT reviewed the data package for the CILRT performed on ; Unit 1, and discussed the conduct of the test with TUEC and NRC ; personnel who participated in or witnessed it. I r K-164 i a
. - - ------m-.y,-g,, - - - , _, ,..~.~---u _,=-wwe--
rvwe-"->-~we- ev *e--- *---*+-w-----=-----e vea-*'**a r-+'--- * - w-wt- ---W--v* v---+= n---*w+--v* --
Apparently after repairing leaks found during the first two attempts, the third attempt at a CILRT was successful. It was successfully completed after three electrical penetrations were isolated because the leakage through them could not be stopped. Tnough the leaks were subsequently repaired and individually tested with satisfactory results, NRC approval was not obtained to perfonn the CILRT with these penetrations isolated. In addition, leak rate calculations were performed using ANSI /ANS 56.8, which is neither endorsed by the NRC nor in accordance with FSAR comitments. Accordingly, TUEC shall identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rather than ANSI N45.4-1972. Additionally, TUEC shall identify to NRC all other deviations from FSAR commitments,
- c. Prerequisite Testing The TRT reviewed FSAR comitments, startup administrative procedures, prerequisite test records, craft personnel qualification records, and discussed them with startup and craft management personnel. The TRT also observed test support craft personnel at work and interviewed soine of them to gain familiarity with their attitudes and capabilities.
The review of test records revealed that craft personnel were signing to verify initial conditions for tests in violation of startup Administrative Procedure-21, entitled: " Conduct of Testing" (CP-SAP-21). This procedure requires this function to be perfonned by System Test Engineers (STE). Startup management had issued a memorandum improperly authorizing craft personnel to perform these verifications on selected tests. Accordingly, TUEC shall rescind the startup memorandum (STM-83084), which was issued in conflict with CP-SAP-21, and ensure that no other memoranda were issued which are in conflict with approved procedures.
- d. Preoperational Testing The TRT assessed the preoperational test program by reviewing administrative procedures, interviewing startup personnel, and examining test records, schedules, system assignments, subsystem definition packages, and the master data base.
Problems found with test data are addressed in section III.a of this enclosure. The TRT also found that STEs were not being provided with current design information on a routine, controlled basis, and had to update their own material when they considered it appropriate. Accordingly, TUEC shall establish measures to provide greater assurance that STEs and other responsible personnel are provided with current controlled design documents and change notices. K-165 l
Attachment 4 , Oceket Nos.: 50-445 00T 5 1!B4 ! and 50-446 Texas Utilities Electric Company Attn: M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
Subject:
September 1B', 1984 Letter, D. G. Eisenhut to M. D. Spence, Re: Comanche Peak Review During our meeting on September 18, 1984 at Bethesda, Maryland, we discussed the technical issues regarding Comanche Peak which the NRC Technical Review Team identified as having potential safety implications and thus requiring
- additional information. The subject letter listing these items and the information that we requested were provided to you during that meeting.
We have since discovered some typographical errors in the Enclosure to the September 18, 1984 letter and provided Mr. John Merritt of your staff with a marked-up copy of that letter on September 21, 1984. Enclosed for your information is an errata to the letter. Sincerely. Original signea by Darrell o,Eisenhat Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosure:
As stated . cc w/ enclosure: See next page i l l
~
l K-167
)
- . . . . _ . . - . = . - _ . _ . . .- . _ . - - - . . - .. ._
Enclosure Errata i To Enclosure 1 to September 18, 1984 Letter, D. G. Eisenhut to M. D. Spence
- 1. Page 2 line 1 Panel CP1-ECPRCB-14 should be Panel CP1-EtPRCB-04
- 2. Page 2, 8th line from bottom of page Panel CPI-EC-PRCP-06 i should be 1 Panel CP1-EC-PRCB-06
]
- 3. Page 4, item 3 Control Panel CP1-EC-PRCP-06
- should be i Control Panel CP1-EC-PRCB-06
- 4. Page 9, 3rd line from bottom of first full paragraph 1
I Sections 3.4.1.1.1 should be Sections 3.8.1.1.1
- 5. Page 10, top of page, item 7 Lowered Suspended Ceiling
, should be Louvered Suspended Ceiling ' K-168 i
CMANCHE PEAK l Mr. M. D. Spence P res'ident Texas Utilities Generating Company 400 N. Olive St. ,- L.B. 81 Dsilas, Texas 75201 l Nicholas S. Reynolds, Esq. Mr. James E. Cummins
. cc: Resident Inspector /Ccmanche Peak i
Sishop, Liberman, Cook, Nuclear Power Station Purcell & Reynolds c/o U. S. Nuclear Regulatory 1200 Seventeenth Street, N. W. l Washington, D. C. 20036 Connission P. O. Box 38 Glen Rose, Texas 76043 l Rcbert A. Wooldridge, Esq. Worsham, Forsythe, Sampels & l Wocidridge y Mr. John T. Collins U. S. NRC, Region IV 2001 Bryan Tower, Suite 2500 611 Ryan Plaza Drive Dallas, Texas 75201 Suite 1000 J Arlington, Texas 76011 i Mr. Hemer C. Schmidt ' Manager - Nuclear Services Mr. Lanny Alan Sinkin Texas Utilities Generating Company 114 W. 7th, Suite 220 Skyway Tower Austin, Texas 78701 400 North Olive Street L. B. 81 B. R. Clements Callas, Texas 75201 Vice President Nuclear Texas Utilities Generating Ccapany Mr. H. R. Rock Skyway Tower Gibbs and Hill, Inc. J 393 Seventh Avenue 400 North Olive Street ,! New York, New York 10001 L. B. 81 Dallas, Texas 75201 ll Mr. A. T. Parker William A. Burchette, Esq. j Westinghouse Electric Corporation 1200 New Hampshire Avenue, N. W. P. 0. Box 355 l Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. . Billie Pirner Garde l Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project 1901 Que Street, N. W. Austin, Texas 78711 Washington, D. C. 20009 l Mrs. Juanita Ellis, President David R. Pigott, Esq. Citizens Association for Sound Orrick, Herrington & Sutcliffe Energy 600 Montgomery Street la26 South Polk San Francisco, California 94111 Dallas, Texas 75224 Ms. Nancy H. Williams ' Anthony 2. Roiiman, Esq. CYGNA Trial Lawyers for Public Justice 2000 P. Street, N. W. ) 101 California Street l San Francisco, California 94111 Suite 611 l l Washington, D. C. 20026 1 ' K-169 l
# "%, Attachment 5 g UNITED STATES I
y n NUCLEAR REGULATORY COMMISSION t;, I WASHINGTON.D.C 20655
***** NOV 2 91984 Docket Nos.: 50-445 and 50-446 Mr. M. D. Spence '
President Texas Utilities Generating Company 400 North Olive Street Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
i
Subject:
Comanche Peak Review On July 9,1984, the staff began an intensive onsite effort to complete a por-tion of the reviews necessary for the staff to reach its decision regarding the licensing of Comanche Peak, Unit 1. The onsite effort covered.a number of areas, including allegations of improper construction practices at the facility. i On September 18, 1984, the NRC met with you and other Texas Utilities Electric Company representatives to provide you with a number of technical issues in the electrical / instrumentation, civil / structural, and test program areas having potential safety implications. The issues discussed constitute a portion of the technical issues and allegations being evaluated by the Technical Review Team (TRT). The activities of the TRT have progressed to the point where it is appropriate to provide you with a status of additional items under review and to request additional information. These items, in the coatings, mechanical, and miscel-laneous areas, are listed in the enclosure to this letter. Further background information regarding these issues will be published in a Supplement to a Safety Evaluation Report (SSER), which will document the TRT's overall assessment of the significance of the issues examined. The items in the enclosure to this letter cover only a portion of the TRT's effort. The TRT's ongoing evaluation, QA/QC review and conversations with allegers may reveal additional items in the coatings, mechanical, and mis-cellaneous areas for which additional requests for information may be appro-priate. Also, the TRT evaluation of QA/QC issues, and its consideration of the programmitic implications of these findings, are still in progress. A summary of these issues will be provided to you at a later date. You are requested to submit additional information to the NRC, in writing, in-cluding a program and schedule for completing a detailed and thorough assess-ment of the issues identified in the enclosure to this letter. This program plan and its implementation will be evaluated by the staff before NRC considers the issuance of an operating license for Conanche Peak, Unit 1. The program plan should address the root cause of each problem identified and its generic K-171 i i
h implications on safety-related systems, programs, or areas. Your You should also address the collective significance of these deficiencies. program plan i should also include the proposed TUEC action to assure that such problems will not occur in the future. This request is submitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding information needs that could potentially affect the safe operation of their plant. Future requests for additional information of this nature will be made, if necessary, as the activities of the TRT progress. Sincerely, j '6f , Darrell GC Eise u - irector
%g/ Division of Licens ngOffice of Nuclear Reactor Ri k
Enclosure:
As stated cc w/ enclosure: See next page I K-172
COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission ' P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels & Wooldridge Mr. Robert D. Martin 2001 Bryan Toker, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin s Skyway Tower 114 W. 7th, Suite 220 Austin, Texas 78701 400 North Olive Street L. B. 81 i Dallas, Texas 75201 B. R. Clements
' Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc. Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker William A. Burchette, Esq.
Westinghouse Electric Corporation 1200 New Hampshire Avenue, N..W. P. O. Box 355 Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C. 20036 Renea Hicks, Esq. Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas 78711 1901 Que Street, N. W. Washington, D. C. 20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq. Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nar.cy H. Williams Anthony Z. Roisman, Esq. CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W. San Francisco, California 94111 Suite 611 Washington, D. C. 20036 i K-173
i COMANCHE PEAK cc: Mr. Dennis Kelley i Resident Inspector - Comanche Peak c/o U. S. NRC P. O. Box 1029 i Granbury. Texas 76048 Mr. John W. Beck Manager - Licensing Texas Utilities Electric Coinpany ' Skyway Tower 400 N. Olive Street L. B. 81 Dallas, Texas 75201
\
Mr. Jack Redding Licensing Texas Utilities Generating Company 4901 Fairmont Avenue Bethesda, Maryland 20014 i i i J . I 1 l t l ' K-174 l l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._______.___________________m___ __ L______
l I I REQUEST FOR ADDITIONAL INFORMATION IV. Protective Coatings Area
- a. Surveillance and Test Program for Coatings The protective coatings Technical Review Team (TRT) reviewed the backfit program, design basis accident qualifications, traceability, application and repair procedures, training, coating exempt log and dispositioning of non-conformance reports. Concurrently, the staff is evaluating the effects on containment emergency sump performance of paint and insulation debris.
The results of the two concurrent reviews will be combined in one supple-mental safety evaluation which is scheduled to be issued by January 1985. Actions required for resolution of protective coatings issues will be delineated in the supplement. V. Mechanical Area
- a. Inspection for Certain Types of Skewed Welds in NF Supports The TRT investigated inspection procedures of Brown & Root (B&R) for welds in pipe supports designed to ASME III Ccde, Subsection NF. The TRT found that no fillet weld inspection criteria existed for certain types of skewed welds. By definition, skewed welds are those welds joining (1) two non-perpendicular or non-colinear structural members, or (2) two members with curved surfaces or curved cross sections, such as a pipe stanchion (a sec-tion of pipe used as a structural member) welded to another pipe stanchion or to a curved pipe pad. Notice that for type (2), the effect of curva-ture at the weld connection induces skewed considerations, even though the two joining members are physically perpendicular. The B&R weld inspection procedures CP-QAP-12.1 and QI-QAP-11.1-28 for NF supports have addressed type (1) skewed welds; however, the TRT found that QI-QAP-11.1-28 did not include weld inspection criteria for type (2) skewed welds. Although the TRT was told by B&R personnel that procedure QI-QAP-11.1-26 for piping weld inspection was used, since such weld connections were similar in con-figuration to a pressure boundary stanchion attachment weld, no evidence documenting the use of this inspection procedure was provided to the TRT.
According to records reviewed by the TRT, these welds were actually cate-gorized as "all other welds" rather than " skewed welds" on the required QC , checklist. Instead of using fillet weld gauges for measuring the size of l nonskewed welds, welders were supposed to use a straight edge and a steel' scale for measurement of a type (2) skewed weld, as described in QI-QAP-11.1-28. In addition, due to the variable profile along its curved weld connection, the weld size should have been measured at several dif-ferent locations. The lack of inspection criteria and lack of verification of proper inspection procedures being conducted for type (2) skewed welds are a violation of ASME Code for NF supports committed to by TUEC in FSAR Section 5.2.1 and a violation of Criterion XVII in Appendix B of 10 CFR 50. K-175 l
- . . -._- ----- _ _ - - . _ . - . _ . - - . . _ ~ - - _ . - - - .
i The TRT reviewed weld inspection procedures, weld data cards, and visually inspected several type (2) skewed welds in randomly sampled NF supports where pipe stanchions were used. Although the small sample of welds inspected by the TRT are acceptable, due to deficiencies in inspection records and the apparent lack of inspection criteria, the TRT is not cer-tain whether other type (2) skewed welds were inspected properly. This is a generic issue involving many NF supports in various safety-related sys-tems. The lack of documented ' inspections and criteria for type (2) skewed l' welds in NF supports represents a safety concern regarding the possible ! existence of under-sized welds in supports which are required to resist various design loads. l Accordingly, TUEC shall I (1) Revise B&R weld inspection procedures CP-QAP-21.1 and QI-QAp-11.1-28 to properly address type (2) skewed welds of stanchion to stanchion and stanchion to pipe pad; and, L (2) provide evidence to verify that previous inspections of these types of skewed welds were performed to the appropriate procedures.
- b. Improper Shortening of Anchor Bolts in Steam Generator Upper Lateral 4 Supports The TRT was informed that some anchor bolts in the steam generator upper i support beams were shortened during installation to less than the length j shown on the design drawing without proper authorization. The TRT was
' told that the bolt cutting incident occurred either because the hole of the anchor device was filled with debris, or the threaded portion of the bolt had concrete mix stuck to it. There are 18 bolts at each end of each of 4 beams, totalling 144 bolts. There is one beam for each steam genera-tor. The bolt threads into an anchor device embedded in the concrete wall.
l The acceptable bolt length or the length of bolt available for threading into the anchor device is vital to ensure structural capability of the support beams. l The TRT attempted to review TUEC records for ultrasonic (UT) measurement results and general installation practices. The TRT was told that ultra-sonic testing of these types of bolts was not a procedural requirement; however, TUEC was' unable to provide any other installation records for TRT review. The TRT concludes that such unauthorized bolt cutting and lack of installation inspection records is a violation of QA procedures and Cri-terion XVII in Appendix B of 10 CFR 50. Since the support beams are essen-l l tial to provide lateral restraint for the steam generator during a LOCA or
; seismic event, adequate anchoring capability M the bolts has safety sig-l nificance and, as a result, appropriate measures are needed to ensure conformance with General Design Criterion 1 of 10 CFR 50.
Accordingly, TUEC shall provide evidence, such as ultrasonic measurement results, to verify acceptable bolt length. Should unauthorized bolt cutting be verified, TUEC shall: K-176 i _._ , _ _ , _ _ , _ , _ . . _ _ _ _ _ . _ , ,_.,,,_.m._ _ _ . _ , _ , , _ _ . _ _ . , . _ . . . _ _ _ _
(1) replace shortened bolts with bolts of proper length, or provide analysis to justify the adequacy of shortened bolts as installed; and, , (2) provide justification or propose measures to ensure that no similar concern exists for bolting.
- c. Design Consideration for Piping Systems Between Seismic Category I and Non-Seismic Category I Buildings In April 1984 the Comanche Peak Special Review Team (SRT), formed and coor-dinated between NRR, IE and Region II and IV, performed a limited review of Comanche Peak. The TRT, in reviewing the SRT findings in the area of piping design considerations, has discovered that piping systems, such as Main Steam, Auxiliary Steam and Feedwater, are routed from the Electrical Control Building (seismic category I) to the Turbine Building (non-seismic category I) without any isolation. To be acceptable, each seismic cate-gory I piping system should be isolated from any non-seismic category I piping system by separation, barrier or constraint.
If isolation is not feasible, then the effect on the seismic category I piping of the failure in the non-seismic category I piping must be considered (CPSES FSAR 3.78.3-13.1). For CPSES, FSAR section 3.78.2.8 establishes that the Turbine Building is a non-seismic category I structure and failure is postulated during the i seismic (SSE) event. The effect of Turbine Building failure on any non-isolated piping routed through the Turbine Building from any seismic category I building must be considered. In addition, for non-seismic category I piping connected to Seismic Category I piping, the dynamic effects of the non-seismic category I piping must be considered in the seismic design of the seismic category I piping and supports, unless TUEC can show that the dynamic effects of the non-seismic category I piping are isolated by anchors or restraints. The anchors or restraints used for isolation purposes must be designed to withstand the combined loading imposed by both the seismic category I and non-seismic category I piping. Accordingly, TUEC shall provide analysis and documentation that the piping systems routed from seismic category I to non-seismic category I buildings meet the stated FSAR criteria. i d. Plug Welds The TRT investigated alleged generic problems regarding uncontrolled repairs to holes existing.in pipe supports, cable tray supports and base
! plates in Units 1 and 2. These holes, which Since had been misdrilled during these support's are Seism #c fabrication, were repaired by plug welds.
i I K-177
1 Category I supports and the effects of the welds have not been evaluated, this constitutes a violation of Criteria IX and XVI of Appendix B to 10 CFR 50. Region IV inspections have confirmed the existence of such welds in cable tray supports located in the Unit 2 Cable Spreading Room. Although the effects of unauthorized, undocumented and uninspected plug welds in some locations (e.g., the webs of I-beams or in structural members in compression) will be inconsequential, their effects in critical loca-tions (e.g., flanges of I-beams in flexure or in structural members in tension) in critically loaded supports or base plates could _ affect their i structural integrity and intended function. i Accordingly, TUEC shall perform one of the following: (1) Modify its proposed plan to Region IV (TXX-4183 and TXX-4259) to include a sampling inspection of all areas of the plant having plug a welds, to include cable tray supports, pipe supports and base plates. Propose alternate methods of inspection where the oblique a' lighting method is not viable (e.g., locations covered by heavy coats of paint). Perform an assessment of the effects on quality due to uncontrolled plug welds found during the proposed inspection, as ,
.I modified above. Submit a report documenting the results of the in-i spection and assessment to the NRC for review.
(2) Perform bounding analyses to assess the generic effects of uncon-trolled plug welds on the ability of pipe supports, cable tray sup-ports and base plates to serve their intended function. Submit a report documenting the results of the assessment to the NRC for review.
- e. Installation of Main Steam Pipes i
The TRT investigated an allegation that a Unit 1 main steam line had been i installed incorrectly and had been forced into proper alignment after flush-ing operations by use of the main polar crane and come-alongs. It was also claimed that pipe supports had been modified to maintain the line in its 4 forced position and vibrations following detachment of-the flushing line could have damaged the main steam line. Based on its investigation, the TRT determined that the alleged incident pertained to restoration of the Unit 1, loop 1 main steam line to its initial, correct installation posi-tion. (The line had_ shifted during flushing operations due to the weight of the added water and because the temporary supports sagged.) The TRT also determined that the modifications to permanent pipe supports were necessary to provide proper support to the main steam line in its restored position (initial designs for and construction of the supports had been 1 based on the shifted position of the line) and, although the alleged vi- ! brations could not be confirmed, their associated stresses might not have 4 i ! damaged the main steam line. (The highest stresses would have occurred _in i the weaker, temporary flushing line.) The TRT~ review of a TUEC analysis, i performed 1 year after the incident, concluded that the analysis was incom-i plete. An evaluation for the full sequence of events leading up to the l i X-178 L-__ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ - _
l 1 incident had not been performed. The TRT review of Gibbs & Hill Specifica-tion No. 2323-MS-100 indicated that there were inadequate requirements and construction practices for the support of the main steam line during flushing, and for temporary supports for piping and equipment in general. In particular, evaluations to assure the adequacy of temporary supports during flushing and installation were not required. The deficiencies in the analyses, specifications and construction practice identified above constitute a violation of Criterion V of Appendix B to 10 CFR 50. Accordingly,'TUEC shall: (1) Modify Gibbs & Hill Specification No. 2323-MS-100, and institu+,e pro-cedures for support of the main steam line during flushing ano for temporary supports for piping and equipment in general to assure that the quality of piping and equipment are not affected. (2) Perform an assessre" of stresses in the portions of the Unit 1, loop 1, main steam and feedwater lines that were affected in the sequence of events involved during their initial installation, flushing and final installation. Conditions requiring stress analysis are: (a) . Flushing condition when the lines were full of water and temporary supports had sagged or settled. (b) Disconnecting condition when vibrations of the temporary line could have occurred. (c) Lifting condition when forces were applied by the polar crane
- and come-alongs.
These assessments shall be based on appropriate piping configurations involved. (3) Perform a non-destructive examination of locations in the Unit 1, loop 1, main steam and feedwater piping where stresses were exceeded during the conditions of concern in a. through c. above. (4) Review the existing baseline UT examinations for those portions of the Unit 1, loop 1, main steam and feedwater involved in all the conditions of concern in a. through c., above, for unacceptable l indications. (5) Review records of hydrostatic testing of the main steam and feedwater line to verify the quality of piping involved in the incident. (6) Provide similar assessments for circumstances involved in a lifting incident identified during the TRT inspection for the Unit 1, loop 4, main steam line. K-179
i (7) Provide assessments of effects on quality of safety-related piping and equipment which were involved in similar incidents of sagging, settlements and failures, if any, of temporary supports. (8) Submit the results of analyses, examinations and reviews in a docu-mented report for NRC review. VI. Miscellaneous Area
- a. Gap Between Reactor Pressure Vessel Reflective Insulation (RPVRI) and the Biological Shield Wall The TRT investigated an allegation that the Unit I reactor pressure
' vessel outer wall was touching the concrete biological shield wall.
A TRT review of existing documentation and discussions with TUEC personnel' indicated that this allegation was not factual. However, a significant construction deficiency report, submitted pursuant to 10 CFR Part 50.55(e), on August 25, 1983, documented that unacceptable cooling occurred in the annulus between the RPVRI and the shield wall during hot functional testing, apparently because of the existence of an inadequately sized annulus gap and possibly because the presence of constre: tion debris in the annulus. TUEC corrected the situation by .rodifications to allow increased air flow for proper heat dissi-pation and by removal of the construction debris. TUEC representa-tives indicated that testing to verify the adequacy of the cooling ficw will take place when additional hot functional testing is con-ducted. Information gathered by the TRT during the investigation indicated that a design change in the RPVRI support ring (i.e., loca-
- ting the ring outside rather than inside the insulation) resulted in a limited clearance between the RPVRI and the shield wall. The TRT review of th'e 50.55(e) report revealed that TUEC failed to
- (1) ad-dress the fundamental issue of the design change impact on annulus
. cooling flow, and (2) determine whether Unit 2 was similarly affected. Accordingly, TUEC shall: 1 (1) Review their procedures for approval of design changes to non- ) nuclear safety-related equipment, such as the RPVRI, and make ! revisions as necessary to assure that such design changes do not adversely affect safety-related systems, t (2) Review procedures for reporting significant design and construc-tion deficiencies, pursuant to 10 CFR Part 50.55(e), and make changes as necessary to assure that complete evaluations are conducted. 1 (3) Provide an analysis which verifies that the cooling flow in the annulus between the RPVRI and the shield wall of Unit 2 is
- adequate for the as-built condition.
K-180 l L - . - . _ . .. - _.- . - - - -_ . . . - . . - - - - . - . _ . . . --
7-(4) Finally, verify during. future Unit I hot functional testing that completed modifications to the RPVRI support ring now allow adequate cooling air flow. The TRT noted that control of debris in critical spaces between components and/or structures was identified as an issue, both in the investigation of this allegation and the civil / structural area item II.c (Maintenance of Air Gap Between Concrete Structures), contained a' in Darrell G. Eisenhut's September 18, 1984, letter to TUEC. Accord-ingly, TUEC shall also: (1) Identify areas in the plant having critical spacing between components and/or st_ructures that are necessary for proper func- , tioning of safety-related components, systems or structures in which unwanted debris may collect and be undetected or be dif-ficult to remove;
, (2) Prior to fuel load, inspect the areas and spaces identified and remove debris; and, ,
(3) Subsequent to fuel load, institute a program to minimize the. collection of debris in critical spaces and periodically inspect th.ese spaces and remove any debris which may be present.
- b. Polar Crane Shimming ,
The TRT investigated the installation of the polar crane rail support system by visual inspection, review of associated documentation, and discussions with TUEC representatives and their contractors. Region IV
.,' Inspection Report 50-445/84-08; 50-446/84-04 and Notice of Violation,
,, dated July 26, 1984, documented that gaps on the Unit 1 polar crane
- bracket and seismic connections exceeded design requirements. In
# Texas Utilitiss Generating. Company responses of August 23, 1984, and September 7, 1984, the gaps were attributed to crane'and bolting
/ .self-adjustment resulting from crane operation. "A site design change 3
(DCA-9872, Rev^ision 4, dated August 24,1984) was issued to document
, the acceptability of the gaps in excess of 1/16 inth which were l identified in the above NRC inspection report.
f e / During further investigation of .theM1egation that shims for the l
- rail support system of the polar cra~ne had been altered during O~ '
~
installation, the TRT observedfgaps which may have been excessive between the crane girder and!the girder support bracket. Detailed specifications addressin6 the gap tolerance in the girder seat con-nections did not exist; however, Gibbs & Hill lettsr GHF-2207, dated l
- November' 28, 1977, stated that the " seated connections will not i require shimming since the area in bearing is at least the width of l the bottom flange of the crane girder." Contrary to this Gibbs & l Hill assumptjon, the TRT observed nine girders with gaps which
.e % a K-181 f .s
l l l extended under the bottom flange that reduced the bearing surface to less than the 20-inch flange width stated in the letter. The TRT also observed conditions which indicated that the crane rail may still be moving in a circumferential direction, that three' rail-to-rail ground wires were broken, that two shims have partially worked out from under the rail, and that two Cadwelds were broken. Accordingly, TUEC shall:
- 1. Inspect the polar crane rail girder seat connections for the presence of gaps which reduce the bearing surface to less than the width of the bottom flange and perform an analysis which will determine whether existing gaps are acceptable or require corrective action.
- 2. Determine if additional rail movement is occurring and, if so, provide an evaluation of safety significance and the need for corrective action.
- 3. Perform a general inspection of the polar crane rail and rail support system, correct identified deficiencies of safety sig-nificance, and provide an assessment of the adequacy of existing maintenance and surveillance programs.
K-182
u 5 Neca A. out Aya , Covo,55so i . . ,0= T Nov i . ,A,.,,, .. r,0 C. , , ,., N. - . ~ N,.,C P w - m NUREG-0797 BIBLIOGRAPHIC DATA SHEET Supplement No. 8
- Le. .e ...
!) TsTLE AND Es5TsTLE 4 RtCPetNT 5 ACCES$ ION NowsER l Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2
- oaf a a Poaf CO PuYi0 MONTM vtAm FEBRUA0Y 1985 6 aLuYMOmt55 i OATE meroaf issuto MONTM vtam FEBDUA9Y 1995 9 PROJECTsT A50wcan UNIT Nuustm 3 6 tasOmuaNG ORGANIZATION Naut AND MAtLING ADORE 55 ##Nt=8e J A Csee#
Division of Licensing "* *""" Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, DC 20555 11 SPON50 ming O.G ANs2 ATION NAME ANO WALLING ADOmg55 gisiewee le Codes las TvPE OP REPOmi TECHNICAL Same as 8. above i2= Psa'oo couaio <d~~ > Julv 1984 Eebruary 1985 13 EuPPLEMENT Amy NOTts Docket Nos. 50-445 and 50-446 la ABSTR ACT (200 swords or asasi Supplement No. 8 to the Safety Evaluation Renort for the Texas Utilities neneratinq Company application for a license to onerate the Comanche Peak Steam Electric Station located in Somervell County, Texas has been . jointly orenared by the M# ice of Nuclear Reactor Regulation and the Technical Review Team of the U. S. Nuclear Requlatorv Commission. This Supplement provides the results of the staff's evaluation and resolution of annroximateh 30 technical concerns and allenations relating to civil / structural and miscellaneous issues regarding construci.fon and olant readiness testino oractices at the Comanche Deak facilitv.
- 5. KEY WOaOS AND DocuwENT ANALv5a5 15e OtsCaiPTOA5 i Ava LAssL Yv ST ATEWENT 17 StCunsiv CLA55sFICA TION lip NuuS6R QF PAGt5 l
Unlimited UfRI'A3fteIED 1 1 5.CuaiTVCLA55.. SCAT,0N re PaiCE I UtittWsi fied s
NUREG-0797 Supplement No. 9 Safety Evaluation Report related t.o the operation of Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446 Texas Utilities Generating Company, et al. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation March 1985 p.= =*a., Qif) jr F0lA-85-59 sf Sll2.(.
_ - . . - - _ - . = _ . . l ! ABSTRACT Supplement 9 to the Safety Evaluation Report for the Texas Utilities Electric l Company's application for a license to operate Comanche Peak Steam Electric Sta-tion, Units 1 and 2 (Docket Nos. 50-445,50-446), located in Somervell County, Texas, has been prepared jointly by the Office of Nuclear Reactor Regulation and the Comanche Peak Technical Review Team of the U.S. Nuclear Regulatory
- Commission. This supplement addresses TUEC's analyses in support of its request
! to amend the Comanche Peak Final Safety Analysis Report to eliminate the commit-ment that coatings inside the reactor Containment Building be qualified for Units 1 and 2. In addition, this supplement provides the results of the staff's evaluation and resolution of 62 technical concerns and allegations in the coat-4 i
- ings area for Unit 1. Because of the favorable resolution of the items discus-sed in this report, the staff concludes for the issues considered herein, that
- there is reasonable assurance that the facility can be operated by TUEC without endangering the health and safety of the public.
I
~
i i i J i 1 I
~
~ 1 Comanche Peak SSER 9 111 l! 1- - - - - _ _ _
TABLE OF CONTENTS . _ Pag _e ABSTRACT........................................................... iii ACRONYMS AND ABBREVIATIONS......................................... vii
- 1. INTRODUCTION.................................................. 1-1
- 2. CONTRIBUTORS TO APPENDIX L.................................... 1-3
- 3. THE COMANCHE PEAK TECHNICAL REVIEW TEAM FOR APPENDIX M........ 1-3 APPENDIX L - The Effects of Paint and Insulation Debris on the Performance of Post-Accident Fluid Systems at Comanche Peak, Units 1 and 2......................... L-1 APPENDIX M - Staff Evaluation and Resolution of Technical Concerns and Allegations in the Area of Protective Coatings at Comanche Peak Steam Electric Station, Unit 1...... M-1 1
Comanche Peak SSER 9 v
4 ACRONYMS AND ABBREVIATIONS AA independent assessment program allegation AB American Bridge AB - bolt allegation ABRR - as-built reverification records A-C - Allis-Chalmers AC - concrete /rebar allegation ACI - American Concrete Institute AD - design of pipe / pipe support allegation ADS - audit discrepancy report AE - electrical allegation AE00 - Office for Analysis and Evaluation of Operational Data (NRC) AFW - auxiliary feedwater system AH - hanger allegation AI - intimidation allegation AISC - American Institute of Steel Construction ALARA- as low as reasonably achievable AM - miscellaneous allegation ANI - authorized nuclear inspector ANS - American Nuclear Society ANSI - American National Standards Institute A0 - protective coating allegation AP pipe and pipe support allegation i APC AMP Product Corporation AQ quality assurance / quality control allegation AQB QA/QC bolt allegation AQC QA/QC concrete /rebar allegation AQE QA/QC electrical allegation
! AQH QA/QC hanger allegation AQL acceptable quality level AQ0 QA/QC coating allegation AQP QA/QC pipe and pipe support allegation AQW QA/QC welding allegation ARMS - Automated Records Management System ASLB - Atomic Safety and Licensing Board ASME - American Society of Mechanical Engineers ASTM - American Society for Testing and Materials AT -
acceptance test AT - test program allegation AV vendor / generic allegation AW - welding allegation ! 1 Comanche Peak SSER 9 vii
B&PVC - Boiler & Pressure Vessel Code B&R - Brown & Root, Inc. BNL - Brookhaven National Laboratory BOC - beginning of cycle BRHL - Brown & Root Hanger Locations BRIR - Brown & Root Inspection Report BRP - Brown & Root piping isometric drawing BTP - Backfit Test Program BWR - boiling water reactor 4 C&L - Corner and Lada (computer program) C&S - civil and structural CAR - Corrective Action Request CASE - Citizens Association for Sound Energy CAT - Construction Appraisal Team (NRC) ] - CB&I - Chicago Bridge & Iron Company CCL - Corporate Consulting and Development Company, Limited
! CCS -
Component Cooling System
- CCW -
component cooling water CEL - Coating Exempt Log CFR - Code of Federal Regulations CHN - construction hold notice CILRT - containment integrated leak rate test 1 CMC - component modification cards CMTR - certified material test report COT - construction operation traveler CP - Comanche Peak CP construction permit CPPE - Comanche Peak Project Engineering CPSES - Comanche Peak Steam Electric Station CPSIG - Comanche Peak Seismic Interaction Group CSS - containment spray system CSTS - Construction and Startup/ Turnover Surveillance Group (TUEC) CVCS - chemical and volume control system CZ Carboline Carbo zinc 11 DBA - design basis accident DCA design change authorization DCC - Document Control Center (TVEC) DCTG - Design Change Tracking Group DCVG - Design Change Verification Group DE - Division of Engineering (NRC)
- DFT -
dry film thickness Division of Licensing (NRC) DL D-6 - Ameron Dimecote 6 Comanche Peak SSER 9 viii 1- . e e s
E&I - Electrical and Instrumentation ECCS - emergency core cooling system E00 - Executive Director for Operations (NRC) E0C - end of cycle E0P - Emergency Operating Procedures ERG - Emergency Response Guideline ETG - Electrical Test Group (TVEC) FOSG - Field Damage Study Group (TUEC) FJO - field job orders FP - fire protection FSAR - Final Safety Analysis Report FW - field weld G&H - Gibbs & Hill GAP - Government Accountability Project GDC - general design criteria GE - General Electric Corporation GED - General Equivalency Diploma Gibbs & Hill hanger (isometric drawing) GHH HFT - hot functional test HIR - hanger inspection rep::rt HP - hanger package HP high pressure HVAC - heating, ventilation and air conditioning system HX - heat exchangers IAP - Independent Assessment Program ICC - inadequate core cooling IE - Office of Inspection and Enforcement (NRC) IEB - Inspection and Enforcement Bulletin IEEE - Institute of Electrical and Electronics Engineers IM - interoffice memorandum (TVEC) INPO - Institute for Nuclear Power Operations IOM - interoffice memorandum IR - inspection report (NRC) IRN - item removal notice ITT-G - ITT Grinnell l JTG - Joint Test Group (TUEC) JUMA - Joint Utility Management Assessment Group LE - left end LOCA - loss of coolant accident LP - liquid penetrant Lomanche Peak SSER 9 ix
., ._ ~ . . - - . ,_
i M&P - mechanical and piping MAR - maintenance action request MCC - motor control center (GE) MDB - master data base MIFI - mechanical fabrication inspector MIL - material identification list (or log) MIME - Mechanical Equipment Inspector MQE Mechanical Quality Engineering MR - material requisition MRS - manufacturer's record sheet MS - main steam (line) MWDC - multiple weld data card N/A - not applicable NCR - nonconformance report (TVEC) NDE - nondestructive examination NDT - nondestructive testing NI - never incorporated NONSAT - nonsatisfactory NOV - Notice of Violation (NRC) NPSH - net positive suction head NPSI - Nuclear Pcwer Service Incorporated NRC - U.S. Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation (NRC) N5SS - nuclear ste:: supply sy3 tem O&M - Operations and Maintenance (TUEC)
, OBE -
operating basis earthquake OI - Office of Investigations (NRC) OJT - on-the-job training OL - operating license ORNL - Oak Ridge National Laboratory PC - protective coating PCR - plant change request PCR - protective coating report PET - permanent equipment transfer PFG - Paper Flow Group PFS - pipe fabrication shop PORV - power operated relief valve PPM - parts per million PSAR - Preliminary Safety Analysis Report PSE - Pipe Support Engineering (TUEC) PT - preoperational test PTS - pressurized thermal shock PWR - pipe whip restraints PWR - pressurized water reactor P-305 - Carboline Phenoline 305 I Comanche Peak SSER 9 x i
i QA quality assurance l QAI quality assurance investigation (TUEC) QC quality control l QE quality engineer RCB - Reactor Containment Building RE - right end RES - Office of Nuclear Regulatory Research (NRC) j RFIC - request for information or clarification (B&R) l RG - Regulatory Guide (NRC) RHRS - residual heat removal system RI - NRC Region I Office RIR - receipt inspection report (TVEC) RIV - NRC Region IV Office RPE - radiation protection engineer RPI - rod position indication RPS - radiation protection supervisor a RPS - report process sheet (TUGCO) RPV - reactor pressure vessel RPVRI - reactor pressure vessel reflective insulation RRI - Resident Reactor Inspector (NRC) RV - reactor vessel RWN - room work notifications i SAP - startup administration procedure I SALP - Systematic Assessment of Licensee Performance (NRC) SAT - satisfactory SAVC - structural assembly verification card SER - Safety Evaluation Report (NRC)
- SG -
steam generator SI - safety injection
; SIS - Special Inspection Services SMAW - shielded metal arc welcing SNM - special nuclear material 50RC - Station Operations Review Committee SRIC - Senior Resident Inspector for Construction (NRC)
SRP Standard Review Plan (NRC) SRT - Special Review Team (NRC) SSE - safe shutdown earthquake 1 SSER - Safety Evaluation Report Supplement SSI - safe shutdown impoundment SSPC - Steel Structures Painting Council SSWP - station service water pumps STE - system test engineer , SWA - startup work authorization SWO - shop work order 4 l Comanche Peak SSER 9 xi
---------,,v .,,---n -
..,e4 ee r awe m n--- , . , ,- v-w.e--,,, -.----...--,..a,--,,- , e-,, -
--.y--,-- e. e y ,e v ---
TOCR - test deficiency change request TOI - Transamerica Delaval, Inc. TDR - test deficiency report 10 CFR 50 - Title 10 Code of Federal Regulations Part 50 TI - temporary instruction TIDC - Division of Technical Information and Document Control (NRC) TNE - TUEC Nuclear Engineering TP - test program TPD - test procedure deviation Tr - transcript TRT - Technical Review Team (NRC) TSABC - technical services as-built coordinator l TSDR - technical services design review coordinator TSI - thermolag TSMO - Technical Services Mechanical Drafting TSP - tri-sodium phosphate TUEC - Texas Utilities Electric Company TUGC0 - Texas Utilities Generating Company TUSI - Texas Utilities Service, Inc. UCC - University Computing Company USI - unresolved safety issue UT - ultrasonic test UTA - University of Texas at Austin , VCD - vendor-certified drawing
- VT -
visual weld (inspector) 1
- W -
Westinghouse Electric Corporation , WDC - weld data card WFML - weld filler metal log WPS - welding procedure specification 4 I Comanche Peak SSER 9 xii
1 INTRODUCTION On July 14, 1981, the U.S. Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) (NUREG-0797) related to the application by the Texas Utilities Electric Company (TUEC) for a license to operate Comanche Peak Steam Electric Station (CPSES), Units 1 and 2. Subsequently, eight Supplemental Safety Evaluation Reports (SSERs) were issued by the staff. This report, Sup-plement No. 9, addresses TUEC's analyses in support of its request to amend the Comanche Peak Final Safety Analysis Report (FSAR) to eliminate the commitment that coatings inside the reactor Containment Building be qualified. Appendix L to this Supplement provides the staff's technical evaluation of the postulated behavior of coatings in the Comanche Peak Containment Building under design basis accident conditions. This evaluation supports TUEC's analyses of its request to amend the FSAR to eliminate the commitment that coatings inside the Containment Building be qualified. Appendix M presents the Technical Review Team evaluation of 62 technical concerns and allegations in the coatings area at Comanche Peak. The actions required in Appendix M were modified based on the staff conclusions in Appendix L. The evaluations for Appendix L and M were performed concurrently by independent groups. Appendix L In the Comanche Peak SER (NUREG-0797), the staff found the coating system inside the Containment Buildings acceptable based on TUEC's commitment in the FSAR to meet the positions of Regulatory Guide 1.54, ANSI N101.2, and ANSI N5.12. Coat-ings which are controlled, applied, and tested to be consistent with these posi-tions are considered " Qualified for a design basis accident (DBA) environment. TUEC proposed to amend the FSAR to eliminate the commitment that coatings inside the Containment Building be qualified on June 4, 1984. In Appendix L, TUEC's proposal was found acceptable, based on a detailed review by the NRC Office of Nuclear Reactor Regulation, which demonstrates that a total failure of protec-tive coatings inside the Containment Building would not adversely affect the performance of post-accident fluid systems. Accordingly, any deficiencies which might result in coating failures would not result in or contribute to causing, or increasing the consequences of, any design basis accident; for this reason, it is not necessary that coatings be qualified. Appendix M The technical concerns and allegations regarding the coatings program at Comanche Peak were part of the outstanding regulatory issues that remained as construction of the Comanche Peak facility neared completion. The NRC's Executive Director for Operaticns (ED0) issued a directive on March 12, 1984, establishing a pro-gram for assuring the overall coordination, integration, and resolution of these issues prior to the staff's licensing decision. In response to the ED0's direc-tive, a program plan was developed and approved on June 5, 1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regu-lation, and the Administrator of NRC's Region IV Office. This program plan, entitled Comanche Peak Plan for the Completion of Outstanding Regulatory Actions, specified the critical path issues, addressed the scope of work needed, and provided a projected schedule for completion. Comanche Peak SSER 9 1-1
Attachment 1 to Appendix M is a listing of the technical concerns and allegations in the protective coatings area. The TRT evaluation of the protective coatings area revealed many specific deficiencies which render a large percentage of the coatings at CPSES unqualified. However, consistent with the findings of Appen-dix L, TUEC has provided justification that debris generated from the failure of all paint in the Containment Buildings under design basis accident conditions will not adversely affect the performance of post-accident fluid systems. Because coatings inside the Containment Buildings need not be qualified, the TRT does not recommend that deficiencies be remedied in the applied coatings at CPSES. Based on TUEC's prior FSAR commitment to provide qualified coatings inside the Containment Buildings, coatings applied before issuance of Appendix L to this SSER were required to have been qualified. The failure of TUEC to fulfill that commitment indicates deficiencies in the quality assurance / quality control (QA/QC) program. Although not of safety significance in the coatings area, these deficiencies will be considered in the TRT's evaluation of the effective-ness of TUEC's overall QA/QC program. Management and coordination of all the outstanding regulatory actions for Comanche Peak are under the overall direction of Mr. Vincent S. Noonan, NRC Comanche Peak Project Director. Mr. Noonan may be contacted by calling 301-492-7903 or by writing to the following address: Mr. Vincent S. Noonan Division of Licensing Office of Nuclear Reactor Regulation U..S. Nuclear Regulatory Commission Washington, D. C. 20555 Copies of this Supplement are available for public inspection at the NRC's Public Document Room at 1717 H Street, NW, Washington, D. C. 20555, and the Local Public Document Room, located at the Somervell County Public Library On The Square, P. O. Box 1417, Glen Rose, Texas, 76043. Availability of all material cited is described on the inside front cover of this report. Comanche Peak SSER 9 1-2 t _ __. .-. _ _ _ . _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _.._.m. _ _ _ _ . _ . . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ - _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ .
- 2. Contributors to Appendix L*
Benaroya, V. Brown, C. - EG&G, San Ramon Burlinger, C. Butler, W. Houston, R. Johnston, W. Kniel, K. Li, C. Lobel, R. Mann, B. Marsh, L. Oliu, W. - Div. of Technical Information and Document Control, NRC Serkiz, A. Shapaker, J. Sheron, B. Sun, S. Witt, F.
- 3. The Comanche Peak Technical Review Team for Appendix M to SER Supplement 9 Brown, C. -
EG&G, San Ramon Gagliardo, J. - Reactor Training Center, IE, NRC Hodgson, B. - Metalweld Inc. Ippolito, T. - Office of Analysis and Evaluation of Operational Data, NRC Johnson, C. - Region IV, NRC Kirslis, S. - Office of Nuclear Reactor Regulation, NRC Lettieri, V. - Brookhaven National Laboratory Matthews, P. - Office of Nuclear Reactor Regulation, NRC McCracken, C. - Office of Nuclear Reactor Regulation, NRC Noonan, V. - Office of Nuclear Reactor Regulation, NRC < Oechsle, S. - Metalweld, Inc. Oliu, W. - Div. of Technical Information and Document Control, NRC Poslusny, C. - Office of Nuclear Reactor Regulation, NRC Tang, R. C. - Office of Nuclear Reactor Regulation, NRC Taylor, J. - Brookhaven National Laboratory Vietti, A. - Office of Nuclear Reactor Regulation, NRC Wells, W. - Metalweld Inc. Wessman, R. - Office of Nuclear Reactor Regulation, NRC Zudans, J. - Office of Inspection and Enforcement, NRC
*Unless otherwise noted, all contributors are from the NRC Office of Nuclear Reactor Regulation.
Comanche Peak SSER 9 1-3
9 = APPENDIX L THE EFFECTS OF PAINT AND INSULATION DEBRIS ON THE PERFORMANCE OF POST-ACCIDENT FLUID SYSTEMS AT COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2
Table of Contents
- P. age
- 1. 0 INTR 000CTION..................................................... L-1 2.0 POTENTIAL EFFECTS OF PAINT DEBRIS ON THE PERFORMANCE OF ENGINEERED SAFEGUARDS............................................ L-1 2.1 Blockage of the Containment Building Emergency Sump Debris Screens..................................................... L-2 2.2 Blockage of Containment Building Spray System Nozzles and RHR/SI System Flow Passages and Equipment........... ... L-6 2.3 Containment Building Hydrogen Generation.................... L-6 2.4 Blockage of Filters in the Containment Building Air Handling Systems............................................ L-6 2.5 Degradation of ECCS Performance by the Entrainment of Fine Particles of Paint 0ebris.............................. L-7 2.6 Fouling of Reactor Core Heat Transfer Surfaces and Core Flow Blockage............................................... L-10
3.0 CONCLUSION
S...................................................... L-13
4.0 REFERENCES
....................................................... L-17
)
I Comanche Peak SSER 9 L-lii
- 1. 0 INTRODUCTION In the NRC Safety Evaluation Report (SER) on operation of Comanche Peak Steam Electric Station (CPSES) (NUREG-0797), the staff found the coating systems to be used inside the Containment Building acceptable based on the Texas Utilities Electric Company (TUEC) commitment in Section 6.1.2 of its Final Safety Analysis Report (FSAR) to apply " qualified" coatings consistent with the positions of Regulatory Guide 1.54 and ANSI N101.2. By letter dated June 4, 1984, TUEC pro-posed to amend the FSAR to eliminate the commitment to apply qualified coatings.
TUEC provided additional information in support of the proposed amendment by letters dated June 29, 1984, July 26, 1984, August 17, 1984, Spetember 10, 1984, October 12, 1984, November 2, 1984, December 17, 1984, January 11, 1985, and February 7, 1985. The NRC Standard Review Plan (SRP), Revision 0 of Section 6.1.2 (November 1975), identified the need to address the quality of paints used inside the reactor Containment Buildings. Revision 1 of SRP Section 6.1.2 (December 1978) addresses the requirements to meet that need by stating that coatings used inside the Con-tainment Building are acceptable if they are applied and tested in accordance with the positions of NRC Regulatory Guide 1.54 and ANSI N101.2, or if the appli-cant provides justification to show that debris generated under design basis accident conditions will not adversely affect the performance of post-accident fluid systems. Revision 2 of SRP Section 6.1.2 (July 1981) retains the position that coatings be applied in accordance with Regulatory Guide 1.54 and ANSI N101.2,, unless the applicant provides an acceptable alternative. The transmittal memo for SRP Section 6.1.2, Revision 0, states that "for some time" after an SRP section is published, NRC will review applications in accordance with prior criteria. NRC staff reviewers were further directed to adapt the SRP section to the particular needs of applications based on positions prior to the latest revision (s) and to make appropriate allowances for the difference in positions when determining their acceptability. Based on the preceding guidance, NRC evaluates each application for compliance with the criteria of SRP Section 6.1.2 consistent with the date the application is docketed and the status of Containment Building painting at the date SRP Section 6.1.2 was issued. Accordingly, the NRC staff has reviewed TUEC's proposed amendment in accordance with Section 6.1.2, Rev. 2, of the Standard Review Plan (NUREG-0800). 2.0 POTENTIAL EFFECTS OF PAINT DEBRIS ON THE PERFORMANCE OF ENGINEERED SAFEGUARDS The design basis accident (DBA) of concern fer coatings in the Containment Building is a loss-of-coolant accident (LOCA) because of the pressure, tempera-ture, and radiation conditions which can exist and which may affect coating adhesion and result in the generation of debris from flaking and peeling. Potential safety concerns stemming from paint debris which could be created from a LOCA in the Containment Building include the following: (a) blockage of Containment Building emergency sump debris screens; (b) blockage of Containment Building spray system nozzles, of residual heat removal / safety injection (RHR/SI) system flow pa', sages and of equipment; Comanche Peak SSER 9 L-1
. . _ . . . - . -. -. -- - - _~ . . _. .
(c) generation of hydrogen in the Containment Building; (d) blockage of filters in Containment Building air handling systems; 1 (e) degradation of emergency core cooling system (ECCS) performance by the entrainment of fine particles of paint debris; and (f) fouling of reactor core heat transfer surfaces and blockage of core flow. Each of these potential safety concerns is discussed in detail in Sections 2.1 through 2.6. 2.] Blockage of Containment Bu_i,lding Emergency Sump Debris Screens TUEC has performed a detailed analysis of the potential for, and the effects of debris blockace of the Containment Building emergency sumps. Generation and transport of both paint and insulation debris were considered in tnis analysis and are discussed in Sections 2.1.1 through 2.1.5. TUEC's analysis follows the guidance and methodology developed by the NRC staff in conjunction I with the work for unresolved safety issues (USI) A-43, Containment Emergency Sump Performance (References 12 and 20). 2.1.1 Paint Debris Effects In its analysis, TUEC postulated that, as a worst case, all coatings inside the Containment Building (the inside surfaces of which are estimated to be 618,000 square feet) could fail and form debris. The transport of such debris following a LOCA was analyzed to estimate the potential sump screen debris blockage and attendant net positive suction head (NPSH) pressure drop (or head loss) for the residual heat removal (RHR) and Containment Building spray pumps. 4 Since the transport of debris is a functior of recirculation water velocity (flowing water with entrained particles) and particle size, TUEC analyzed the containment flow fields and the entrainment characteristics as a function of particle size. These analyses showed that paint debris would not be transported i from far regions (relative to sump location) within the Containment Building
- due to the very low recirculation velocities predicted. This part of the analysis, termed "far-field" effect, supported reducing the region of concern to the 60'-0-315 azimuthal area. This effect allowed for reducing the amount of failed paint which could potentially affect screen blockage to approximately 95,400 square feet.
)
For the coated surfaces in the immediate vicinity of'the sump screens, where-flaking paint particles could fall directly into the pool near the screens, TUEC performed a "near-field" analysis of paint particles to assess potential screen blockage. The model calculated the trajectory of the particles in the pool as a function of local water velocity and settling velocity to determine screen blockage. These two-dimensional flow field calculations showed that due to the presence of the sump cover plate overhang, an area of at least 24 square feet would remain open at the top of the sump screen structure (each sump has a i total fine screen area of 356 square feet), resulting in a maximum estimated j blockage of approximately 95 percent, regardless of particle size. Particles less than 0.125 inches (1/8 inch) in diameter which did not settle out on~ lower screen regions would pass through the debris screen. L-2 Comanche Peak SSER 9
i TUEC estimated a blocked screen head loss of about 0.4 feet of water, using data obtained from full-scale, hydrau'lic model testing of the Comanche Peak ! sump design at Western Canada Laboratory; independent NRC staff assessments of the same experimental data estimated the head loss to be 0.5 0.3 feet of water. Since the available NPSH margin for the RHR pumps and containment spray pumps is 4.23 feet and 5.81 feet of water, respectively, adequate NPSH margin will remain despite paint debris blockage. The staff reviewed the analytical approach used by TUEC in applying the calcula-tional methods and findings from NUREG/CR-2791 (Reference 20) and NUREG-0897, Revision 1,* to analyze potential sump debris blockage caused by paint debris transport and the associated impact on NPSH margin. Based on this review and on independent assessments, the staff finds that TUEC's analysis is sufficiently conservative and that the postulated failure of all Containment Building coatings would not substantially reduce the NPSH margin to the RHR and CSS pumps. The related details of TUEC's analysis and staff findings are presented in the sections which follow. 2.1.1.1 Characteristics of Paint Debris Approximately 285,000 square feet of concrete and 333,000 square feet of steel surfaces in the Comanche Peak Containment Building are covered by coatings which have a specific gravity ranging from 1.5 to 4.0. As a worst case, TUEC assumed that all coatings failed completely. The failure modes for coatings are chalking (powdering) and peeling (flaking). TUEC assumed the latter failure 4 mode occurred, which is the more conservative assumption from the standpoint of its potential for sump screen blockage. The coatings applied at CPSES consisted primarily of inorganic zinc primers and organic top coats of the same generic type as have been qualified by other nuclear plants for the design basis accident (DBA) environment. Exposure of these coatings, if properly applied, to DBA conditions of temperature, pressure and radiation does not result in flaking or peeling. If improperly applied, enough flaking and/or peeling may occur at DBA conditions to generate debris. TUEC's assumption that all coatings inside of the Containment Building will i fail and generate debris is conservative. With inorganic zinc coatings, a slight amount of " chalking" occurs, creating powder particles in the micron size range. Organic coatings remain intact, showing at most a few small blisters. The fine powder particles from the chalking of inorganic zinc coatings are inert and would have no tendency to adhere to each other or to solid surfaces. Inorganic coatings suffer relatively , little radiation damage compared to organic coatings. For the generic types of organic coatings used in the CPSES Containment Build-ing, the degree of cross-linking and disolvation (loss of plasticizer and mon-omer) would be increased by heat and radiatinn exposure, thus making the polymers harder and more brittle (References 1 and 2). Flakes of this nature would not agglomerate or stick to other solid surfaces. Some coatings may flake off Containment Building surfaces because they were not properly applied. ,
*NUREG-0897, Rev. 1 (Unpublished), " Containment Emergency Sump Performance--
Technical Findings Related to Unresolved Safety Issue A-43," March 30, 1984. Comanche Peak SSER 9 L-3 l l u , _ _ . _ __ ____ __. _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ __ _
Since the fine screens of the Containment Building sumps have 1/8-inch open-ings, larger debris has the potential to block the sumps. TUEC conservatively assumed 1/8-inch particles in the sump blockage analysis, since these would lead i to screen blockage, and because larger debris particles are more difficult to transport to the vicinity of the sump screens due to the generally low flow velocities predicted to occur in the Containment Building following a LOCA. i The effects of paint debris smaller than 1/8 inch are addressed in Sections 2.2, 2.4, 2.5 and 2.6 below. 2.1.2 Flow Field Water Velocities
< The flow pattern within a Containment Building is plant-specific and is a function of the building's interior design. For the Comanche Peak Containment Building, spray headers are located in the dome and at each floor elevation.
Tc calculate one-dimensional global water velocities in the Comanche Peak
< Containment Building during the recirculation phase, TUEC applied methodology developed as part of the staff resolution of USI A-43, as reported in NUREG/
CR-2791 (Reference 20). More detailed two-dimensional flow field calculations were performed to evaluate local debris blockage effects. The water velocity at a given point in the building determines the transport potential for debris. The calculated water velocities were compared with the threshold velocity needed to transport debris of a selected size and were found to be too low to transport paint debris from far regions of the Containment Building. The NRC staff reviewed TUEC's approach to applying the USI A-43 debris transport methodology and evaluated the uncertainties and conservatisms in calculating water velocities. The acceptability of TUEC's characterization of certain plant-specific features which could have an important effect on the outcome of the analysis were verified by the staff during a site visit. Based on its review, the staff finds that the water velocities have been calculated in a reasonable manner for the debris transport evaluation. 2.1.3 Paint Debris Transport
- The paint debris transport model was based on the methods described in NUREG/
CR-2791 (Reference 20) for the transport of insulation debris. TUEC studied , the transport of paint particles of different sizes and densities, and deter-mined that a threshold water velocity of 0.27 ft/sec would be required to tran-sport 1/8-inch paint particles. Since water velocities on the upper floors range from 0.3 to 1.2 ft/sec, all paint debris generated on the upper floors could be ' transported to the sump level. (Four-inch curbing would direct debris toward stairways and grates.) At the sump level, the calculated water velocities are generally below 0.2 ft/sec; however, at doorways and other reduced area passage-ways, velocities on the crder of 0.4 to 0.7 ft/sec were calculated. Although general water velocities in the pool on the containment floor were calculated to be less than 0.27 ft/sec, TUEC assumed that all debris near the sump zone could be transported to the sump because of the increased velocities at reduced area passageways. That is, only 95,400 square feet out of 618,000 square feet of coating debris is available to be transported to the containment floor at the sump level. TUEC conservatively assumed that the paint particles would be transported, accumulating at the base of the screens with a 45-degree angle of Comanche Peak SSER 9 L-4
I i repose for blockage estimates. The threshold water velocity of 0.27 ft/sec for ! 1/8-inch paint particles calculated from TUEC's far-field debris transport model
- is conservative compared with the velocity considered necessary to suspend and i
transport coal slurries of similar particle size and density. Water velocities in excess of 2 ft/sec are required to transport 1/8-inch coal particles (specific
- gravity = 1.4) in either a suspended condition or as a sliding bed in large diameter horizontal pipes (References 3, 4, and 5). At lower velocities, the particles tend to settle out. The viscosity of water at a temperature of 200 F L is 1/3 its value at room temperature. Therefore, the resistance to particle settling is decreased, and the threshold velocity needed for transporting 1
particles would be correspondingly increased. The threshold velocity may also be estimated from the observation that "a
- linear velocity equal to about seven times the terminal [ settling] velocity of the largest particle in the slurry is necessary to maintain adequate suspension"
.) (Reference 4). The terminal settling velocity for coal particles pas;ing a 1 1/16-inch sieve is given as 0.224 ft/sec (Reference 5). Coal particles of this
! size would weigh approximately the same as 1/8-inch diameter, 0.005-inch-thick paint discs. The water velocity for adequate suspension in horizontal flow at , ambient temperature is, therefore, about 1.57 ft/sec. At higher temperatures, higher water velocities would be needed for particle transport because of
] decreased viscosity and density. Particles would move by saltation or as a
! sliding bed at a somewhat lower velocity. By either estimation, TUEC's value of 0.27 ft/sec for threshold water velocity is judged to be conservative.
2.1.4 Near Sump Effects Results of TUEC's study of near sump effects indicate that a large area of the , fine screens can be blocked by paint debris, assuming that paint fragments j larger than the minimum screen opening reach the sump, stick to the surface, and do not overlay other fragments. A solid steel plate at the top of the sump extends more than 1 foot outward from the fine screens, and approximately j 8 inches outward from the coarse screen. The near-sump fragment trajectory . j analysis and detailed two-dimensional flow field calculation show that a 2-inch ?' band at the top of the fine screen will remain free of paint debris due to this design feature. In addition, a portion of the screen facing away from the pri-3 mary sources of paint debris will remain partially open. , TUEC's transport model for the region near the sump is also conservative in ' considering the terminal settling velocity of paint chips to be 0.16 f t/sec for the tumbling chips. The data for equivalent 1/16-inch coal particles indicated 1 a settling velocity of 0.224 ft/sec at ambient temperature and approximately i three times this value at 200 F (Reference 6). Based on the staff's independent i evaluation of TUEC's debris transport model, there is reasonable assurance that { significant sump screen blockage by paint debris will not occur. 2.1.5 Insulation Oebris The generation and transport of insulation debris were analyzed in accordance i with the guidance presented in NUREG-0897 (" Containment Emergency Sump Perform-i ance"), Rev. 1. There are three types of insulation used inside the Comanche Peak Containment Building. The first is the reflective metallic type, which comprises the largest portion of insulation. The second type is high-efficiency j Comanche Peak SSER 9 L-5 i i
metallic thermal insulation, composed of fibrous media and fine, heat-resit. ant particulate matter totally encased in stainless steel. The third type is an antisweat insulation used on cold water piping, which is fiberglass, e'ncapsu-lated in metal casing. Since jet impingement from pipe breaks is the most significant debris generation mechanism for insulation, the analysis considered 20 high energy pipe break locations to maximize estimated debris generation. For reflective metallic insulation, the hot-leg break in steam generator (SG) compartment 4, which is the closest to the sumps, was selected and evaluated for debris generation. For reflective metallic and high efficiency metallic insulation, flow velocities along the flow pathways are insufficient to transport them to the sump. For the third type, encapsulated insulation, none of the postulated pipe breaks in ' the vicinity of this type of insulation were sufficiently energetic to cause actuation of the safety injection or containment spray systems. The recircula-tion flow velocity analysis provided showed that transport velocities between the " break" compartment and the sump were less than 0.2 ft/hr; therefore, transport of insulation debris initially deposited in the SG compartment is highly unlikely. Therefore, the staff has concluded that insulation debris will not contribute to sump blockage. 2.2 Blockage of Containmen Building Spray _ System Nozzles and RHR/SI System Flow Passages and Equipment . The recirculation inlets to the RHR/SI and containment spray systems are pro-tected by sump structures composed of trash racks, coarse screens, and fine screens. The fine screens are sized to preclude particles larger than 1/8 inch in diameter from passing through. The containment spray nozzles, spray pumps, and RHR/SI system flow passages can accommodate 1/8-inch particles without clog-ging. Therefore, the staff finds that the fine screens of the sump structures will protect against the injection of debris which could have a detrimental effect on the emergency core cooling and containment spray systems. 2.3 Containment Building Hydrogen Generation The primer coat on steel surfaces is a zinc-based paint which, on exposure to hot water, can oxidize to form zinc oxide and release hydrogen gas from the water. The design basis hydrogen generation analysis for the plant was previ-ously calculated based on the assumption that all zinc in the coatings reacts to form hydrogen. Therefore, because the maximum generation has already been taken into account, there is no need to consider hydrogen generation any further. 2.4 Blockage of Filters in the Containment Building Air Handling Systems Under LOCA conditions, it is not likely that paint debris would become air-borne, because of the scrubbing effect of the containment spray and the high density of the debris. The only system having filters which could potentially be affected by airborne paint chips is the post-accident containment atmosphere cleanup system. However, since the system is a small, low-capacity system, and would only be operated in the long term following onset of an accident, at which time any potential airborne paint debris would have settled, it would not pose a problem relative to system performance. Comanche Peak SSER 9 L-6
I ; I
- 2. 5 Degradation of ECCS Performance by the Entrainment of Fine Particles of i Paint Debris 4
! Sections 3.2, 3.2.2, and the Appendix of a Gibbs & Hill Report (Reference 7) discuss the potentially adverse effects to emergency core cooling system (ECCS) performance following a complete failure of the Containment Building paint during a design base LOCA in which paint particles of 1/8 inch or smaller pass i through the containment sump screens. The staff requested additional informa-1 tion regarding: (1) the long-term corrosion and erosion effects on ECCS l performance and heat transfer capability, and (2) calculation of the extent and consequences of core flow blockage by paint flakes of 1/8 inch or smaller, i Subsequently, the staff requested additional clarification of TUEC's responses on the large break LOCA and information on the effects of containment paint i failures for a spectrum of small break LOCAs. TUEC's responses to these { requests and to Reference 7 are evaluated below. 2.5.1 Large Break LOCA With respect to blockage of the ECCS flowpaths due to entrained paint parti-
! cles, TUEC provided additional information in References 7 through 10. TUEC performed a bounding analysis in which it was assumed that approximately 278 1
1 cubic feet of paint (i.e. , the paint volume in the 60*-0-315 sector of the Containment Building above and adjacent to the ECCS sumps) would be entrained in the ECCS fluid; the resulting total debris concentration suspended in the , circulating fluid would be less than 1 percent by volume. TUEC indicated that I too types of regions within the ECCS recirculation flow paths may make it 1 possible for paint particles to settle out of the fluid streams, i.e. , regions j in which the fluid flows vertically at low velocity (such as the reactor vessel 1 lower plenum), and long horizontal pipes in which the fluid velocities are low ' ] I enough that the paint particles settle out at the bottom. TUEC also indicated t h t for small pipes and heat exchanger tubes with higher flow velocity, the -{ paint particles would be carr'ad along with the fluid. 2 i To evaluate TUEC's findings that the ECCS would not be blocked, the staff re-
] viewed the flow paths in the low head recirculation mode and concluded that i
there are no small valves or small bore orifices that would be blocked due to l the entrained paint particles. With respect to blockage and fouling due to l agglomeration on the surfaces in the ECCS low head recirculation flow path, the i RHR heat exchanger tubes would be most limiting. Blockage and fouling of the i tubes would not occur under the combined temperature and radiation conditions that would prevail during the LOCA recirculation phase due to the physical ! properties of the paint, as discussed in Section 2.1.1.1, above. 3 The staff has generically evaluated the effects of debris entrainment of up to
' 2.5 volume percent en RHR pump performance (Reference 12). In Reference 12 the staff recommends an assessment-of RHR pump peformance based on plant specific '
estimated debris concentrations and properties. TUEC estimated that the maximum
} concentration of entrained debris would-be less than 1 percent (References 7 ; through 10). The staff independently verified TUEC's estimated debris concen- ;
- tration. In Reference 10, the RHR pump vendor states that the concentration
, and properties of debris estimated by TUEC are expected to have a negligible j effect on pump performance. Based on the preceding evaluation, reasonable assurance exists that the RHR pumps will not be adversely affected by failure i of all coatings in the Containment Building. j Comanche Peak SSER 9 L-7 l l
4 l TUEC has also indicated in Reference 10 that the potential for clogging the pump seal cavity and recirculation tubing is very low since the seal cavity is isolated from the main pump flow by close clearance between the pump shaft ring bushing and the pump shaft. The staff concurs with this conclusion based on information contained in Reference 12, which evaluates RHR pump performance in scenarios similar to this situation. The staff concludes that TUEC has provid-ed reasonable assurance that the low head recirculation portion of the ECCS would retain its heat removal capability under these conditions. 2.5.2 Small Break LOCA In the event of a small break LOCA larger than 1 inch, or a stuck open power j operated relief valve (PORV), high head recirculation could be utilized, requiring operation of the safety injection and/or centrifugal charging pumps, and utilizing the high head injection lines, which are relatively small and contain orifices and throttling valves. With regard to the latter, TUEC indi-j cated that the high fluid velocities (25 to 40 ft/sec) in these lines, valves, r and orifices preclude any line clogging from paint particles. Based on a staff request, TUEC obtained the orifice diameters and flow areas of the valves in
; their throttled condition and found that the clearances are sufficient. Also, ,
there is sufficient flow path redundancy to assure that SI flow to the RCS I would be retained even if the smallest clearance throttle valve became clogged. ! The staff expressed its concerns about the effect of sespended paint particles on sustained operation of the multi-stage high head SI pumps, which would be more sensitive to entrained solids than the single stage RHR pumps, and about the effect of RCP restart on core flow. The staff request for additional information on these subjects was as follows: (1) The staff cannot conclude from the information provided for small break LOCAs that the high head pumps can perform acceptably in the presence of paint particles in the sump coolant. In order to complete our review, the applicant must dem.nstrate that either: i (a) For all design basis accidents, recirculation from the sump using the i high head pumps is not required to meet any of the Commission's i regulations. In particular, for the small break LOCA, the applicant must demonstrate that the long term cooling requirement of 10 CFR 50.46 is met without reliance on the high head pumps in the recircu-lation mode. If demonstration relies on operator action, we require j evidence that the operator can reliably perform the necessary ac-
- tions. This evidence must include a demonstration that the operators have had sufficient training in the necessary action,.have demon-strated their ability to perform the necessary action, (e.g., on a i simulator), and that the emergency operating procedures (EOPs) have been revised to clearly provide the necessary guidance, or, I (b) Provide evidence, preferably experimental or test data, to show that j the high head pumps can perform acceptably over the period of time
! necessary to maintain long term core cooling assuming paint particles j Comanche Peak SSER 9 L-8
l ! in the sump water. As a minimum, we require a statement from the pump manufacturer regarding their pumps' performance capability in this mode. (2) Discuss whether the reactor coolant pumps could be restarted during the recirculation phase and the potential for core blockage assuming the 4 maximum debris volume stored in the reactor vessel lower plenum. l l TUEC responded (Reference 11) that after a small-break LOCA, the total volume ! of debris passing through the sump screen would be less than 10 cubic feet. i Such a volume would result in a total debris concentration in the circulating i coolant of less than 200 ppm, of which less than half would be abrasive. The methodology for calculating particle transport utilized in this reference l was evaluated in Sections 2.1.3 and 2.1.4 of this Appendix and found to be conservative. TUEC further assumed that the percentage of fines below 1/8 inch in diameter in the paint debris was not over 1 percent for concrete paint and i not over 5 percent for steel coatings. The staff independently confirmed that these estimates of the size distribution of failed coating particles were con-sistent with DBA test experience and paint industry experience. TUEC's calcula-tions of fine debris volume considered only the 95,400 square feet of paint in 1 the 315 60 azimuthal region of the Containment Building. TUEC's previous j conclusion that paint debris from the remaining azimuthal regions would not j reach the sump was based on the assumption that particles would be 1/8-inch in
; diameter. If the fine particles from all 618,000 square feet of coatings could i reach the sump, TUEC's det, ailed method of calculation (Reference 11) would . yield a total fine debris of approximately 20 cubic feet. However, TUEC's j analyses in supplementary Table 2 of Reference 11 demonstrate that only fines i of less than approximately 4.0 mils could be transported from far field regions j because flow velocities at the sump elevation, from these regions to the sump screens in the 60 315 azimuthal Containment Building region are, in most cases, less than 0.1 feet per second. Based on the staff's independent analyses and on the information in References 3, 4 and 5 on particle transport, most of the particles outside the 60 315 azimuth would settle before reach-i ing the sump screens. Those that reached the sump would amount to only a frac-
- tion of the total fines.
i Therefore, the staff finds that TUEC's estimates of fine debris volume, and of 4 the concentrations of abrasive and non-abrasive debris concentrations in the 3 circulating coolant, are acceptably conservative. ) TUEC described operability tests that have been performed by the manufacturer on multi-stage high head and low head safety injection pumps with a debris concentration of 92 ppm, including 60 ppm of abrasive materials (concrete and l gl a's s ). The pumps were operated in the hot condition for a period of 10 hours
! (Reference 11) and the change in hydraulic performance was less than 1 percent.
l Vibration levels did not change throughout the tests, and when the pumps were
! disassembled there was no sign of wear, nor did the mechanical seal exhibit i leakage before, during, or after the test. The CPSES pumps are of similar i design to those tested, and utilize the same wear-resistant materials for the impellers, wear rings, and seals. The CPSES pump vendor, Pacific Pumps, has 1 provided a letter (attached to Reference 11) that states that the CPSES charging l j and SI pumps would " function properly without any significant impairment in performance" with a total of 200 ppm of debris, including 100 ppm abrasives.
Comanche Peak SSER 9 L-9
Based on the information above, including the pump vendor's statement regarding operability, the staff concludes that TUEC has provided reasonable assurance that the safety injection portion of the ECCS would retain its operability with the postulated debris concentration. The staff was concerned that restart of the RCPs could introduce paint particles of up to 0.125 inches into the reactor core if the ECCS were operated in the recirculation mode, since the lower plenum fluid velocity at restart would be sufficient to carry particles previously trapped in the lower plenum into the core. With respect to the question regarding RCP restart following recircula-tion, the staff determined that the RCPs could be restarted under the following conditions: (1) During the post-LOCA cooldown and depressurization; (2) In the event of inadequate core cooling (ICC), if the core exit thermocouples indicate a temperature equal to or greater than 1200 F; (3) In the event of an imminent pressurized thermal shock (PTS) condition. In Reference 13, TUEC indicated that RCP restart during ICC would have an event frequency of less than 10 8 per reactor year. The staff considers this number optimistic, but concurs with TUEC that restart under these conditions has a low probability. The staff concludes that RCP restart in the recircula-tion mode because of PTS is also a low probability event. During the post-LOCA cooldown and depressurization mode, the Emergency Response Guidelines (ERGS) instruct the operator to restart the RCPs if both generic and plant-specific restart criteria are met. TUEC indicates that RCP restart during the post-LOCA cooldown for breaks large enough to require recirculation is not likely because of Comanche Peak plant-specific requirements (e.g., restart of component cool-ing flow and leak-off flow if terminated as a result of containment isolation). Additionally, in Reference 19 TUEC stated that the attachments to the Emergency Response Guidelines (ERGS) would be changed to state that a RCP should not be started if containment spray has been actuated and transfer to cold leg recir-culation has been performed, with the exception of ERG FRC-0.1 (Response to Inadequate Core Cooling). As discussed above, inadequate core cooling is a low probability event. The staff concludes that RCP restart following ECCS recirculation is unlikely at Comanche Peak. The staff finds that its and TUEC's independent analyses provide reasonable assurance that in the event of a large or small break LOCA, coincident with LOCA-induced Containment Building paint failure, the long term decay heat removal capability of the ECCS would not be unacceptably degraded. 2.6 Fouling of Reactor Core Heat Transfer Surfaces _and Core Flow Blockage The staff reviewed TUEC's information in References 8, 9, and 10 to assess core flow blockage caused by paint particles detaching from the Containment Building surfaces during a large break LOCA. Based on this review, the staff concluded that because the particles carried into the reactor core are smaller than the minimum flow area in a fuel bundle and can therefore pass through the core, core flow blockage caused by these paint particles is not a concern during cold leg recirculation following a large break LOCA. Subsequently, in Comanche Peak SSER 9 L-10
4 l response to NRC questions, TUEC indicated in Reference 11 that under certain conditions the operator is required to restart the RCPs following a small break LOCA. Restart of the RCPs would result in larger particles being carried into the core from the reactor vessel lower plenum. The staff was concerned that larger particles could accumulate in the core and cause flow blockage. The results of the evaluation of this concern are given below. 2.6.1 Large Break LOCA For a postulated large break LOCA, it is assumed that paint particles are strip-ped from Containment Building surfaces and transported into the reactor vessel during ECCS recirculation. Whether the paint particles would settle to the bottom of the reactor vessel lower plenum or would be transported into the core can be calculated by performing a force balance on a particle of a given size 4 which accounts for gravity,. drag, and hydrostatic forces. The important
- parameters to determine the particle settlement are fluid velocity and particle
! size. TUEC indicated in Reference 8, and the staff independently verified, that the maximum fluid velocity at the entrance to the core during ECCS recirculation following a large break LOCA is 0.3 ft/sec, assuming a hot leg break with two j RHR pumps operating. Given this fluid velocity, TUEC calculated the largest i particle size that could enter the core during ECCS recirculation to be less than 0.036 inches in diameter, which is smaller than the minimum dimension of l 0.040 inches in the spacer grids. As a result of this calculation, TUEC con- ! cluded that the particles carried into the core will pass through the core, and thus will not block flow in the fuel assemblies. i Particles up to 0.125 inch in diameter can pass through the sump screens and I are potentially available, but the calculation described above shows that the i low flow velocity will not carry these particles into the core. Larger parti-cles transported into the vessel will settle out in the reactor vessel lower plenum. These particles are brittle and do not stick at the temperature exist-i ing during ECCS rec rculation. The largest particles carried into the core i are smaller than the minimum flow area in the core. In view of these considera-
; tions, the staff agrees with TUEC that the particles transported into the core will pass through it and not block flow during ECCS recirculation following a 3
large break LOCA.
! 2.6.2 Small Break LOCA As concluded in Section 2.5.2 of this Appendix, RCP restart following ECCS recirculation is unlikely at'CPSES. This section evaluates-the potential for core blockage in the unlikely event that an RCP is restarted.
4 For ECCS recirculation with restart of reactor coolant pumps (RCPs).following i a small break LOCA, it is assumed that the high fluid velocity will result in particles up to the 0.125-inch maximum being carried into the core. i Reference 11 estimates that less than 10 cubic feet of paint particles can be-carried into the vessel during this scenario. A simple, bounding calculation, assuming that all materials are carried into the core and are stopped by the
- . first grid, shows that a layer up to 1-inch thick would form. Such a layer j could result in flow blockage, but is not expected to have this effect for i
several reasons. First, as stated previously, the paint particles would remain brittle and would not adhere to the grids or fuel rods. Second, any significant ! Comanche Peak SSER 9 L-11 1 1 L ___-____ __ _-____-_ __ __-__ _ _ _ ___ _ _ __-__ ______- __-___ _ __-_-___-_-- _ -_________._-___________---
~ _ . - . ._
~
A I -
- accumulation would cause flow pressur6 gradients which would tend to disperse the smaller particles before the buildup could become widespread, and therefore significant in terms of core cooling. The pressure difference could also break It is difficult, up the brittle,sthin paint particles into smaller particles.
however, to determine how the effects of pressure gradient would be balarced with the-accumulatio&of particles in the small spaces between the grids and fuel rods. Even if paint particles were to accumulate, the information provided in Refer-ence 16 demonstrates that a porous blocksge (such as that most likely to be formed in the case of paint particle accumulation) would still provide "a small residual flow," which would result in a flow pattern +hich is "on the whole... nearly unchanged" from the unblocked flow pattern, Based on this result, fuel failure sould not be expected. Even if flow blockage were to occur, the fuel bundie aesign at CPSES is such that the ratio of heat transfer area to flow area is large so that a large fraction of the flow area must be blocked before there is a significant heatup of the fuel rods. In Reference 18, it is con-cluded that for a boiling water reactor (BWR) fuel assembly, a flow reduction of greater that 79 percent is necessary to cause loss of nucleate boiling and that a flow blockage of greater than 95 percent is necessary for fuel cladding to melt. While these numbers include a small amount of bypass flow which is available in a BWR assembly when the inlet is blocked, the numbers provide a qualitative indication that the amount of blockage required to produce these adverse effects is large. A BWR fuel assembly is surrounded by a channel box which does not permit flow to merge with the rest of the core. The pressurized water reactor (PWR) fuel assemblies to be used at Comanche Peak, with their open lattice configuration, would be expected to decrease the consequences of a blockage of only one assembly due to cross-flow from adjacent assemblies. The analyses of Reference 18 were done assuming full core flow. At Comanche Peak, only one RCP would be in operation following a small break LOCA. The lower core ficw could tend to make these results less conservative, but the qualita-tive results would still be valid. The Comanche Peak FSAR reports that tests on simulated PWR fuel bundles of the type used at CPSES (Reference 15) indicate that blockages of 41 percent of the subchannels in the center of a bundle between spacer grids will not cause significant loss of flow because the stagnation zone disappears a short distance past the blockages. The FSAR concludes that " local flow blockages within a fuel assembly have little effect on subchannel enthalpy rise." The staff concurs with this conclusion and has cited these data in the past in connection with concerns over local blockages caused by loose parts. The results if an extrapolation of the BWR calculations and PWR bundle data to a core-wide blockage is not clear and has not been addressed by TUEC. , The staff's qualitative conclusion based on-the facts presented above is that the flow blcckage must be extensive in order to cause fuel rod damage. Restart of a RCP following ECCS recirculation is unlikely. For the reasons discussed above, a complete blockage at the lower fuel assembly grid is unlikely if an RCP is restarted. A flow blockage sufficient to'cause fuel failure is also unlikely. If localized flow blockage were to occur, the staff would expect the extent of fuel failure, if any, to be low. q '
~
j N O ' Comanche Peak SSER 9 L-12 u e .
i The staff finds that TUEC's analyses provide reasonable assurance that in the event of a large or small break LOCA that is coincident with LOCA-induced con-tainment paint failure, the long-term decay heat removai capability of the ! ECCS would not be unacceptability degraded. The staff also concludes that RCP restart following a small break LOCA is unlikely If an RCP were to be restarted following recirculation, a complete blockage at the lower fuel assembly grids due to paint particles is unlikely. A flow blockage sufficient to cause fuel failure is also unlikely. If localized flow blockage were to occur, the extent of fuel failure, if any, would be expected to be low. L Long-term core coolability would not be impaired. 2.6.3 Reactivity Effects of Paint Particles Following either a small or large break LOCA, there is a possibility that paint debris could enter the water circulating through the ECCS via the centainment sump. TUEC has estimated the amount of paint debris that could be present in the circulating water for both the large break LOCA (in amounts of less than 1 percent by volume) and the small break LCCA (in amounts of about 200 parts ; per million). (The paint debris consists mainly of zinc primer and various organic compounds.) The staff evaluates, below, the reactivity effect of TUEC's estimates of paint debris for the reactor at end of cycle (E0C) when, following a small or a large break LOCA, the circulating water would contain about 1800 ppm of boron.* (The results for beginning of cycle [BOC] are similar.) This amount of boron would shut down the reactor. Even assuming that no control rods were inserted, the reactor would be shut down by no less than approximately 900 ppm of boron. Although the worth of boron varies with the reactor state, the staff will assume a 1 percent reactivity worth per 100 ppm of boron. Consequently, the reactor effective multiplication factor, K would be much less than 0.95 for these conditions. Making the conservatiNf,ssumption a that the paint debris would cause a positive reactivity effect by diluting the circu-lating borated water by 1 percent with unborated water, the reactivity effect (approximately 0.2 percent reactivity) would be significantly less than that required for criticality. The staff concludes, therefore, that the reactivity effect of the paint debris in the circulating borated water would be of no concern for criticality following either a small or a large break LOCA.
3.0 CONCLUSION
S Based on the preceding evaluation, the staff f_inds reasonable assurance that debris generated by the failure of all coatings inside the. Containment Building under design basis accident conditions will not unacceptably degrade the perfor-mance of post-accident fluid systems. The failure of coatings would not result in or contribute to causing, or increasing the consequences of, any design basis accident. Accordingly, such coatings are not required to meet the standards of 10 CFR Part 50, Appendix B. Therefore, TUEC's proposal to amend the FSAR to eliminate the commitment to apply qualified coatings is approved. Although the ability to achieve safe shutdown and to maintain long-term core cooling is not degraded by the failure of all coatings inside of the reactor Containment Building, the maintenance of a quality coatings system is beneficial. ,
*From a mixture of primary system (0 ppm boron) and refueling water storage tank (2000 ppm boron).
Comanche Peak SSER 9 L-13
4.1 Preoperational 4.1.1 Testing of coatings as applied to demonstrate their ability to retain adequate adhesion under a range of operating conditions. This should include in situ temperature and pressure testing with separate evaluation and consideration of the effects of radiation exposure, and concurrent adhesion testing in directly adjacent areas. 4.1.2 Complete and careful visual inspection, using optical aids, such as binoculars, of coated surfaces to detect current or e incipient failures. Temporary scaffolding should be used selectively in areas of particular interest. 4.1.3 Information on failure characteristics from in-situ tempera-ture and pressure tests which fail should be assessed to ensure that such characteristics do not adversely affect post-accident fluid systems performance. 4.2 Post Operations 4.2.1 Complete visual inspection, as in 4.1.2 above, at each refuel-ing outage. 4.2.2 Repetition of testing as in 4.1.1 and 4.1.3 at every third refueling outage to detect the capacity of coatings to withstand DBA conditions over time. This inspection should include the use of temporary scaffolding and lighting which is crected and utilized consistent with ALARA guidelines for workers perform-ing these tasks. 5.0 General Requirements for Testing and Surveillance 5.1 Locations of testing and surveillance should be selected so as to provide special attention to Containment Building areas closest to the sumps. 5.2 Specific emphasis should be given to coatings areas which are in the coatings exempt log. Comanche Peak SSER 9 L-16
GUIDELINES FOR A PRE- AND POST-0PERATIONAL C0ATINGS TESTING AND SURVEILLANCE PROGRAM FOR COMANCHE PEAK UNITS 1 AND 2
- 1. 0 Purpose 1.1 The program is intended to provide information on the ability of the applied coatings to reasonably maintain their integrity without sepa-rating from the surfaces to which they have been applied.
2.0 Responsibilities 2.1 TUEC shall propose a program which addresses all the criteria set forth herein, and shall establish procedures for the implementation and documentation of that program. 2.2 The NRC will review and approve the program prior to its implementation. 3.0 Method 3.1 TUEC shall develop comprehensive written instructions that describe how coating testing, surveillance, and repairs will be performed. 3.2 These instructions should include the following: 3.2.1 The qualifications and training of the personnel who implement the program to ensure good workmanship. 3.2.2 The tests, surveillance, and repair procedures which will be implemented. 3.2.3 Detailed operational methods for each test, surveillance, and repair procedtre to ensure good workmanship. 3.2.4 Instruments and apparatus to be employed and the accuracy requirements, calibration method, and calibration frequency for each. 3.2.5 Frequency of tests and surveillance routines, both in terms of sample sizes and scheduling of repeated testing and surveillance. 3.2.6 Acceptance criteria for each test, surveillance, and repair activity. 3.2.7 Records to be maintained to document all of the above. 4.0 Testing and Surveillance The program should include the following: l Comanche Peak SSER 9 L-15 i
-' l 1
l _ l
l l l 4.1 Preoperational 4.1.1 Testing of coatings as applied to demonstrate their ability to retain adequate adhesion under a range of operating' conditions. This should include in-situ temperature and pressure testing with separate evaluation and consideration of the effects of radiation exposure, and concurrent adhesion testing in directly adjacent areas. 4.1.2 Complete and careful visual inspection, using optical aids, such as binoculars, of coated surfaces to detect current or incipient failures. Temporary scaffolding should be used selectively in areas of particular interest. 4.1.3 Information on failure characteristics from in-situ tempera-ture and pressure tests which fail should be assessed to ' ensure that such characteristics do not adversely affect post-accident fluid systems performance. 4.2 Post Operations 4.2.1 Complete visual inspection, as in 4.1.2 above, at each refuel-ing outage. 4.2.2 Repetition of testing as in 4.1.1 and 4.1.3 at every third refueling outage to detect the capacity of coatings to withstand DBA conditions over time. This inspection should include the use of temporary scaffolding and lighting which is erected and utilized consistent with ALARA guidelines for workers perform-ing these tasks. 5.0 General Requirements for Te ting and Surveillance 5.1 Locations of testing and surveillance should be selected so as to provide special attention to Containment Building areas closest to the sumps. 5.2 Specific emphasis should be given to coatings areas which are in the coatings exempt log. i f i Comanche Peak SSER 9 L-16
._ . . . _ _ . _ _ _ _ _ _ _ . _ . . . _ . - m_____ _ _______.__ _ _ . - . - __._._ _ _ _ ____ .____-__- _ _ _ _ _ .___- - . _ _ _ ___._ _ .
4.0 REFERENCES
- 1. Bolt, R. O. and J. G. Carroll, Radiation Effects on Organic Materials, Academic Press, New York, 1963 Chapter 12,
- 2. Parkinson, W. W. and 0. Sisman, "The Use of Plastics and Elastomers in Nuclear Radiation," Nuclear Engineering and Design 17 (1971), pp. 247-280, North-Holland Publishing Co., Amsterdam.
- 3. Worster, R. C. and D. F. Denny, Proc. Inst. Mech. Engrs. (London), 16_9, 563-586 (1955)
- 4. Durand, R. and E. Condolios, " Hydraulic Transport of Coal and Other Solid Materials in Pipes," National Coal Board, Great Britain, Nov. 5-6, 1952.
- 5. Newitt, D. N., et al., Trans. Inst. Chem. Engrs., 33, 93-110 (1955).
- 6. Lammers, G. C., et al., U.S. Bus. Mines Rept. Invest. 5404, 1958.
- 7. Gibbs & Hill Report, " Evaluation of Paint and Insulation Debris Effects on Containment Emergency Sump Performance," June 1984.
- 8. Letter from H. C. Schmidt, TUGCO, to B. J. Youngblood, NRC, " Containment Sump Performance," July 26, 1984.
- 9. Letter and report from J. W. Beck, TUGCO, to B. J. Youngblood, NRC,
" Evaluation of Paint and Insulation Debris Effects on Containment Emergen-cy Sump Performance," Revision 1, October 1984.
- 10. Letter from W. J. Beck, TUGCO, to B. J. Youngblood, NRC, " Evaluation of Paint and Insulation Debris-Effects on Containment Energy Sump Perfor-mance," November 2, 1984.
- 11. Letter from W. J. Beck, TUGCO, to B. J. Youngblood, NRC, " Evaluation of Paint and Insulation Debris-Effects on Containment Energy Sump Performance," December 17, 1984.
- 12. NUREG/CR-2792. "An Assessment of Residual Heat Removal and Containment Spray Sump Performance Under Air and Debris Ingestion Conditions," Septem-ber 1982.
- 13. Letter from R. S. Howard, Westinghouse, to J. T. Merritt, TUGCO, "CPSES Containment Paint Evaluation," January 8,1985.
- 14. Letter from W. J. Beck, TUGCO, to B. J. Youngblood, NRC, " Containment Sump Performance," January 11, 1985.
- 15. P. Basmer et al., Atomwirtschaft, dated August 1972, P416.
- 16. D. Kirsch, " Investigations on the Flow and Temperature Distribution Downstream of Local Coolant Blockages in Rod Bundle Assemblies", Nuclear Engineering and Design 31 (1974) PP266-273.
Comanche Peak SSER 9 L-17
l
- 17. Tang, Y. S., et al., " Thermal Analysis of Liquid Metal Fast Breeder Reactors," American Nuclear Society, 1978. ;
- 18. " Consequences of a Postulated Flow Blockage Accident in a Boiling Water Reactor," GE Topical Report NED0 10174, Rev. 1, October, 1977.
- 19. Letter from W. J. Beck, TUGCO, to B. J. Youngblood, NRC, February 7, 1985.
- 20. NUREG/CR-2791. Wysocki, J. J. et al. , " Methodology for Evaluation of Insulation Debris," September 1982.
Comanche Peak SSER 9 L-18 w
xs - - w - --- + - a f 44 e i APPENDIX M NRC STAFF EVALUATION AND RESOLUTION OF TECHNICAL CONCERNS 1 AND ALLEGATIONS REGARDING PROTECTIVE C0ATINGS INSIDE OF THE REACTOR CONTAINMENT BUILDING AT COMANCHE PEAK STEAM ELECTRIC STATION UNIT 1 '1 1 I
, - - . . - . . - , .-.. , ,,r~.....,n,.. - , . - . . , - , . - - , , , , , , .,,-, . -- - ,, - , , - ,. .
TABLE OF CONTENTS P_ag
- 1. Introduction................................................. M-1
- 2. Comanche Peak Technical Concerns and Allegations Management Program.................................................... M-3 2.1 Background.............................................. M-3 2.2 Review Approach and Methodology......................... M-3 2.2.1 Concern and Allegation Tracking System........... M-3 2.2.2 Review Methodology............................... M-4 2.2.3 Interviews with Allegers......................... M-5 2.3 Communications with TUEC................................ M-5
- 3. S umma ry o f Ev a l ua ti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . M-7 3.1 Scope of Concerns and A11egations....................... M-7 3.2 Protective Coatings Group............................... M-8 3.3 Findings for Protective Coatings Issues................. M-9 3.4 Overall Assessment and Conclusions...................... M-10
- 4. Actions Required of TUEC..................................... M-13 4.1 Backfit Test Program.................................... M-13 4.2 Traceability............................................ M-13 4.3 Coatings Procedures..................................... M-13 4.4 Coatings Exempt Log..................................... M-14 Attachments 1- Listing of Technical Concerns and Allegations in the Coatings Area......................................................... M-15 2- Assessment of Individual Technical Concerns and Allegations in the Coatings Area......................................... M-23 l
l Comanche Peak SSER 9 M-iii l l
T 1
- 1. Introduction As construction of the Comanche Peak Steam Electric Station was nearing comple-tion, issues that remained to be resolved prior to the consideration of issu-ance of an operating license were complex, resource intensive, and spanned more
. than one NRC office. To ensure the overall coordination and integration of these issues, and to ensure their resolution prior to licensing decisions, the NRC's Executive Director for Operations (EDO) issued a memorandum on March 12, 1984, directing the NRC's Office of Nuclear Reactor Regulation to manage all necessary NRC actions leading to prompt licensing decisions, and assigning the 'j Director, NRC's Division of Licensing, the lead responsibility for coordinating and integrating the related efforts of various offices within the NRC.
- The principal areas needing resolution before a licensing decision on Comanche Peak can be reached include: (1) the completion and documentation of the staff's review of the Final Safety Analysis Report (FSAR); (2) those issues in
) contention before the NRC's Atomic Safety and Licensing Board (ASLB); (3) the
- completion of necessary NRC regional inspection actions; and (4) the completion d and documentation of the staff's review of technical concerns and allegations j regarding design and construction of the plant.
1 1 Technical concerns and allegations about Comanche Peak, totalling approximately 900, have been raised mainly by the quality assurance / quality control (QA/QC) personnel working or having worked on site. Their job responsibilities. involve or involved QA/QC aspects of safety related structures, systems, and components to determine whether and to what extent such items are manufactured, purchased, j stored, maintained, installed, tested, and inspected as required by project documents and procedures. Many of these allegations were made orally to NRC Region IV staff, NRC Comanche Peak Site Resident Inspectors, NRC investigators, or in letters to the NRC, as well as in testimony before the Atomic ~ Safety and Licensing Board (ASLB). Individuals with allegations were also sponsored by the intervenor group Citizens Association for Sound Energy (CASE)'and the Government Accountability Project (GAP). General allegations about poor con-struction work at Comanche Peak were also made in several newspaper articles
- in the Dallas / Fort Worth, Texas areas.
By the end of April 1984, the staff identified approximately 400 technical . concerns and allegations related to the construction of the Comanche Peak i facility, including findings by NRC's Special Review Team.- (See Section 2.1 i below.) During its investigation of a concern or' allegation, the TRT identi-fied additional concerns. Interviews with allegers also yielded additional concerns. By December 1984, approximately 600' concerns and allegations had , been identified. In addition, approximately 300 allegations were recently ! provided to the staff by one alleger, bringing the total of concerns and alle-gations to approximately 900. t
- These technical concerns and allegations were grouped by subject into the following areas:
l- Comanche Peak SSER 9 M-1
1 i i Electrical and Instrumentation Civil and Structural Mechanical and Piping 4 - Quality Assurance and Quality Control (QA/QC) Coatings Test Program Miscellaneous This report covers protective coatings inside the reactor Containment Building and is the third of a series of reports dealing with the NRC staff's efforts to evaluate and resolve the technical concerns and allegations raised by various parties and individuals regarding the Comanche Peak facility. Reports on the staff's evaluation of technical concerns and allegations in electrical and instrumentation and test programs (SSER 7), and in civil and structural and miscellaneous areas (SSER 8), were published in January and February 1985, respectively. An allegation or concern was determined to be without safety significance if, based on technical findings, the assessment showed that a structure, component, or system would perform its intended function. The technical concerns and allegations in the areas of mechanical and piping and QA/QC, as well as the remaining areas of outstanding regulatory actions, will be addressed in future supplements to the Comanche Peak Safety Evaluation Report (SER). The staff's findings for coatings allegations and concerns are summarized in Section 3 of this Appendix. Attachment 1 to the appendix is a listing of coatings concerns and allegations about Comanche Peak. Details.of the assess-ment and findings on individual concerns or allegations appear in Attachment 2 to this Appendix. Those aspects of the concerns or allegations that pertain to wrongdoing (e.g. , falsification of records) were forwarded to the NRC's Office of Investigations (OI) for followup because they are outside the scope of the technical staff's review. A number of potential violations of NRC rules and regulations have been identi-fied during the course of the TRT investigation. These potential violations have not been addressed in this SSER, but will be reviewed further by the NRC Region IV staff, which will determine appropriate followup actions. Comanche Peak SSER 9 M-2
- 2. Comanche Peak Technical Concerns and Allegations Management Program
2.1 Background
Shortly after the ED0's issuance of the March 12, 1984, directive, the staff found it necessary to (1) obtain current information relative to TUEC's manage-ment control of its construction, inspection, and test program and (2) obtain necessary information to establish a management plan for resolution of all outstanding licensing actions. In order to achieve these goals, a Special Review Team (SRT) was formed to conduct an unannounced review of the Comanche Peak plant. The SRT consisted of eight reviewers and one team leader, all from NRC's Region II Office, and a team manager from NRC headquarters. The SRT spent over 800 man-hours, from April 3 to April 13, 1984, performing this re-view. The SRT concluded that TUEC's programs were being sufficiently controlled to allow continued plant construction while the NRC completed its review and inspection of the Comanche Peak facility. The SRT review also provided a basis for the development of an NRC management plan for the resolution of all outstanding licensing actions. This plan was approved on June 5, 1984, by the Directors of NRC's Office of Inspection and Enforcement, Office of Nuclear Reactor Regulation, and the Administrator of NRC's Region IV Office. The purpose of the plan was to ensure the overall coordination and integration of the outstanding regulatory actions at Comanche Peak and their satisfactory resolution prior to a licensing decision by the NRC. In accordance with the plan, a Technical Review Team (TRT) was formed to evaluate and resolve technical issues and those allegations that had been identified. On July 9,1984, the TRT began its 10-week (five 2-week sessions) onsite effort, including interviews of allegers and TUEC personnel, to deter-mine the validity of the technical concerns and allegations, to evaluate their safety significance, and to assess their generic implications. The TRT con-sisted of approximately 50 technical specialists from NRC headquarters and Regional Offices, as well as NRC consultants. TRT members were divided into groups according to technical discipline. Each group was also assigned a group leader. 2.2 Review Approach and Methodology 2.2.1 Concern and Allegation Tracking System A tracking system was developed for identifying and listing each concern or allegation. These technical concerns and allegations were grouped according to their topical areas or disciplines, and were listed numerically within each group in the order that they were identified by the TRT. The tracking system included a description of the concern or allegation; its status or the actions taken to resolve it; the nature of the sources of the concern or allegation (i.e. , anonymous or confidential); a code for the individual who identified the concern or allegation (instead of the individual's name); the date the concern or allegation was received by the TRT; the source document (e.g., letter, NRC inspection report, hearing transcript, etc.); and cross reference. At the end of each 2-week session, the concern / allegation tracking system was updated, as needed, to reflect the status of each concern or allegation, as well as any new ones that had been added. Comanche Peak SSER 9 M-3
l l 2.2.2 Review Methodology The technical concerns or allegations similar in subject were combined and evaluated as one category. For each concern / allegation or concern / allegation category, an approach to resolution was developed by the cognizant reviewer (s). Each approach to resolution was reviewed and approved by the responsible group leader. The group leaders and reviewers were instructed to: develop and maintain a work package for each issue or category of issues that contained or referenced pertinent documentation associated with the issue (s) and the ultimate resolution, including records of interviews and inspections for supporting the final NRC staff decisions regarding the issue (s); and to protect the identity of.the allegers, as is the NRC's practice. Such ef- I forts included limited and controlled distribution of allegation-related documents; minimal use of names, identifying titles, or position descrip-tions in written material; enlarged sampling of activities to prevent direct links by non-NRC personnel between the activity under investigation and the alleger; and other indirect approaches toward investigating the allegations. During TRT onsite sessions, daily meetings were held at the review group level to assess progress, to adjust the inspection and evaluation approach as needed, and to provide a forum for the reviewers to interact with one another or to discuss problems and to arrive jointly at resolutions. Similar daily meetings were also held at the management level where the group leaders interacted with one another and with the Project Director, his. assistant and staff. In evaluating the technical concerns and allegations, the TRT reviewers exam-ined areas in the plant where direct observation could provide information needed for evaluating an allegation or concern. During its onsite sessions, the TRT interviewed the allegers as needed to clarify their concerns or allega-tions. To the extent possible, the TRT contacted allegers after its onsite review to discuss preliminary TRT findings and to obtain any additional com-ments from them. (See Section 2.2.3 below.) The TRT also interviewed TUEC and TUEC contractor personnel as was warranted by the evaluation. In addition to these contacts, the TRT reviewed various project documents, including specifi-cations, engineering drawings and analyses, procedures, instructions, NRC Region IV inspection reports, and applicable sections of the Final Safety Analysis Report (FSAR) and NRC regulations pertinent to the allegation or sample selected by the TRT for inspection. The TRT also examined construction records, such as design change authorizations, construction work packages, QC inspection reports, nonconformance reports, deficiency logs, lists and reports, and QC inspector training and certification records. In addition, the TRT reviewed pertinent transcripts from recent ASLB hearings, other sworn testimony of TUEC personnel and former employees, and reports from ?'RC's Office of Inves-tigations (0I). Based on these reviews and interviews, the TRT determined the validity of each technical concern or allegation and assessed its safety significance and its potential generic implications. Detailed documentation of the TRT assessment and final determinations of each technical concern or allegation appear in Attachment 2 to this Appendix. Comanche Peak SSER 9 M-4
2.2.3 Interviews with Allegers In January 1984, RIV contracted with Brookhaven National Laboratory (BNL) to provide technical assistance for onsite reviews and technical evaluation of the allegations of deficiencies related to the protective coatings program at CPSES. On April 25, 1984, BNL sent an interim report on protective coatings to RIV. This repurt requested information from TUEC regarding the Backfit Test Program (BTP), presented BNL's independent test results, and reported BNL's interim findings on TUEC's protective coatings procedures and documentation. On June 13, 1984, BNL sent a draft " Status Report on Protective Coatings Alle-gations" to RIV. This report provided the current status of BNL's investiga-tion of 60 allegations regarding protective coatings at CPSES. On July 9, 1984, the TRT assumed responsibility for completing the investigation of protective coatings allegations. By memo dated August 7, 1984, from D. Eisenhut to ASLB, the status of the 60 allegations under investigation by BNL were provided to the ASLB. These 60 allegations, plus two additional allegations, formed the basis for the coatings TRT investigation. The TRT Coatings Group reviewed 62 allegations made by 12 allegers. Many of these allegations were similar or identical; therefore, most allegations were identified by more than one alleger. During its onsite work, the TRT Coatings Group interviewed 3 allegers who were associated with 57 of the 62 allegations to obtain additional information about the issues involved. Attempts to interview individuals associated with the remaining 5 allegations were unsuccessful. Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original concerns and to obtain any additional comments from them. (Each of these interviews was transcribed.) Eight allegers participated in followup interviews, two declined followup interviews, and two could not be found. Seven of the eight allegers who participated in the followup interviews indicated that they were satisfied that the TRT had investigated the concerns. The eighth alleger indicated that although the TRT had investigated most of the concerns, this alleger was not satisfied that allegation AQ0-36. which addressed the dispositioning of an NCR, was sufficiently investigated. In the opinion of the TRT Coatings Group, the generic evaluation of coatings traceability (Category 3) encompassed the concerns expressed by this individual. The TRT Coatings Group substantiated 23 allegations, partially substantiated 20 allegations, and turned over 3 allegations, which involved intimidation or wrongdoing, to the Office of Investigations. Sixteen allegations were not substantiated. 2.3 Communications with TUEC Whenever the TRT reviewers encountered problems during their evaluations, the TRT Project Director and/or his designee resolved them through discussions with TUEC management onsite. There were also frequent staff-level contacts between TRT members and TUEC personnel during the TRT's onsite activities. In keeping with the NRC practice of promptly notifying applicants of information/ evaluation needs that could potentially affect plant safety, the staff held several meetings with TUEC representatives at NRC headquarters toward the end of the TRT's review. These meetings were held to discuss potential safety concerns and to request additional information needed by the TRT to complete its review. Comanche Peak SSER 9 M-5
_ . - _ - - - - . - - . .- - - _ - . - _ - . _-. . ~ J d i ! i i The NRC staff met with TUEC representatives for the first of these meetings on September 18, 1984, to discuss TRT findings for electrical and instrumentation, i civil and structural, and test program allegations and concerns. A letter documenting these findings and a request for additional information was issued to TUEC on the day of the meeting. TUEC.later submitted the requested informa-tion in the form of a proposed program plan, delineating planned actions to address the deficiencies identified by the TRT. The TRT met with TUEC represen-tatives to discuss this proposed program plan on October 19 and 23, 1984. TUEC submitted a partially revised program plan to NRC on November 21, 1984. On November 29, 1984, NRC sent a letter to TUEC containing potential open issues and requesting additional information and proposed program plans for mechanical and piping and miscellaneous allegations and concerns. The letter also advised TUEC of the status of NRC's evaluation of coatings allegations. On January 8, 1985, the NRC issued a letter to TUEC informing them of the TRT's preliminary findings in the construction QA/QC area and requesting a program and schedule for completing a detailed and thorough assessment of the QA issues presented in the letter. A meeting between TUEC and the TRT was held on January 17, 1985, to discuss potential open issues in the QA/QC area. TUEC's proposed program plan for each of the subject areas and its implementation of the plan will be evaluated by the NRC staff prior to the NRC licensing decision on Comanche Peak. 4 I i l t t Comanche Peak SSER 9 M-6 l i I
1 1
- 3. Summary of Evaluations 3.1 Scope of Concerns and Allegations The concerns and allegations in the Protective Coatings area relate to all important aspects of pre-construction and construction activity, including qualification and traceability of coating materials; procedures for surface preparation, application, and inspection; training and qualification of person-nel; identification and resolution of deficiencies; the backfit test program; documentation; and improper management pressure upon inspectors. Sixty-two allegations in the coatings area were received and evaluated by the TRT, and
- each was assigned, as appropriate, to one or more of seven general categories of concern established by the TRT.
In each of the seven general categories, the TRT Protective Coatings Group con-ducted a generic review of the area of concern, as well as a specific investi-gation and evaluation of each allegation assigned to the category. The seven categories and descriptions of their associated concerns follow: Category Characterization of Concerns Number Subject and Allegations 1 Backfit Test Inspections for coatings adhesion, Program thickness, and visible defects were not performed properly. Inspection procedures were inadequate. Inspection results were j not properly evaluated and documentbd. 2 Design Basis Some protective coating systems applied at Accident (DBA) CPSES were not DBA qualified. Qualification Testing 3 Traceability The traceability of coating materials was not always maintained. Coating materials were not properly stored. Coating materials were contaminated.
- ~4 Coatings Procedures permitted the use of unqualified c Procedures coating systems. Procedures included instructions which were technically
! incorrect. Backfit inspection procedures and methods were inadequate. Coatings were applied to surfaces where they should not have been ap' plied. QC inspections were j inadequate. Procedures were inadequate to
, assure traceability, i Comanche Peak SSER 9 M-7
Category Number Subject Characterization of Concerns and Allegations 5 Inspection Deficiencies were not properly identified, Reports, evaluated, resolved, and documented. Nonconformance Reports, and Design Change Authorizations 6 Coatings Exempt No specific allegations were associated with Log the coatings exempt log. Instead, the TRT conducted a generic review due to its concerns with the size of the total exempted area and the adequacy of methods used to identify and document exempt items. 7 Training and Training and qualification of inspectors and Qualification painters were inadequate and not properly documented. Some personnel were not qualified for their assigned tasks. The TRT did not investigate allegations or concerns on issues of improper manage-ment pressure, intimidation, harassment, or wrongdoing. Allegations AQ0-16, AQO-56 and AQ0-60 dealt with those concerns and were not included in any of the seven categories evaluated by the TRT. Allegation AQ0-16 and AQO-56 involved intimi-dation of coatings QA inspectors by a TUEC civil QC supervisor. These allega-tions, which were investigated by NRC Office of Investigations (01) and reported in 01 Report 4-83-001, dated August 24, 1983, are currently before the CPSES ASLB. Allegation AQO-60 involved selective management assignment of certain QC inspectors so that coatings work would pass inspection. As part of its inves-tigation of other technically related allegations, the TRT found that TUEC maintained a log that identified the coatings QC inspector work assignments and the plant area to be inspected. While there were indications of occasional assignment changes, the TRT could make no conclusion of wrongdoing without further investigation. By letter dated August 24, 1984, the TRT Director forwarded the transcribed NRC interview with this alleger to the NRC OI for their review. Many of the allegations applied to more than one category. Attachment 1 pro-vides a listing of each allegation and the category or categories in which the allegation is reviewed. 3.2 Protective Coatings (PC) Group The PC Group which performed the onsite investigation and assessment of the allegations consisted of six reviewers, each of whom has had experience with protective coatings in nuclear power plants. Three are associated with a national laboratory; two of the three are consultants with contracting and con-sulting experience in this area. Two team members from the NRC Office of j Nuclear Reactor Regulation (NRR) have experience reviewing protective coatings ! Comanche Peak SSER 9 M-8
in nuclear plants. The seventh reviewer was an inspector from the NRC Region IV office. (Unless otherwise noted, the PC Group will be referred to as the TRT in the remainder of this report.) 3.3 Findings for Protective Coatings Issues In the Comanche Peak Safety Evaluation Report (NUREG-0797), issued in July 1981, the coating system inside of the Containment Buildings was found to be accept-able based on TUEC's commitment in its FSAR, Section 6.1.2, to meet the positions of Regulatory Guide 1.54, ANSI N101.2, and ANSI N5.12. Coatings which are con-trolled, applied, and tested to be consistent with these positions are considered to be " Qualified" for a design basis accident (DBA) environment. On June 4, 1984, TUEC proposed to amend the FSAR to eliminate the commitment that coatings inside the Containment Building be qualified. In Appendix L to this supplement, TUEC's proposal to eliminate this commitment that coatings be qualified was found to be acceptable, based on data and analyses which demonstrate that a total failure of protective coatings inside both Containment Buildings would not adversely affect the performance of post-accident fluid systems. Consequently, the staff agreed with TUEC that coating failures do not have safety significance. Therefore, the staff accepted TUEC's position that qualification of these coat-ings snould no longer be required at CPSES. However, based on TUEC's prior FSAR commitment to provide qualified coatings inside of Containment Buildings, coatings applied before issuance of Appendix L of this supplement were required to have been qualified. TUEC's failure to fulfill that prior commitment indicates deficiencies in the coatings QA/QC pro-gram. These deficiencies, although not of safety significance in the coatings area, will be considered in the TRT's overall evaluation of the effectiveness of TUEC's quality assurance / quality control (QA/QC) program. The TRT finds that all of the allegations except the one related to the backfit test program were not of significant technical concern, either because they were not substantiated or because they involved relatively small areas of coating. The one exception pertains to coatings on miscellaneous steel items which failed the adhesion test. These areas may amount to 6 percent of the total area coated. The generic review and some of the substantiated allegations demonstrated pro-cedural and implementation deficiencies in quality assurance and quality control during the Backfit Test Program. As a result of deficiencies in TUEC's documentation, design, and engineering functions, the TRT finds that TUEC has not demonstrated that the coating sys-tems applied at CPSES are DBA qualified. In addition to this shortcoming, a number of allegations involving relatively small areas were substantiated in which improper coating sequences or procedures were permitted without adequate , engineeringjustification. The TRT finds that TUEC's inadequate inspection and documentation practices for coating work prior to November 1981, resulted in loss of material traceability. The Backfit Program tests provide an indication of the quality of the tested coatings, but are not a substitute for coating traceability. The TRT's generic review and evaluation of individual allegations related to coating procedures has led to the finding that, in many cases, the procedures l Comanche Peak SSER 9 M-9
were inadequate and resulted in applied coating systems which were either not qualified or not technically viable. These cases include special coating sys-tems for overlaps and repairs, incorrect power tool cleaning instructions, and inadequate instructions for the " nickel test," for final reinspection of repair work, for illumination during visual inspection, for the proper use and testing of compressed air for spray painting, for masking areas that should not be coated, and for maintaining traceability of coating materials. These proce-dural deficiencies indicate inadequate performance by those responsible for the review and approval of the coating procedures. The TRT finds that the specific allegations pertaining to the disposition of unsatisfactory inspection reports, nonconformance reports, and design change authorizations are not of significant technical concern, either because they were not substantiated, or because the concerns were resolved properly. How-ever, in its generic review, the TRT finds that, in many cases, nonconformance reports and design change authorizations were dispositioned without documenta-tion of adequate engineering evaluation and justification. Although there were no specific allegations dealing with the Coatings Exempt Log (CEL), which is a record of unqualified coatings inside the Containment Building, the TRT conducted a generic review of the log because it provided a convenient measure of the total area of plant coatings with unacceptable or indeterminate quality. For these items, which were listed in the CEL, the TRT finds that the determination to include them was made in a conservative manner - and the method of estimating the item area was reasonably conservative. However, several sizable areas with coatings of indeterminate quality, for example approximately 54,000 square feet of coatings which may have failed the adhesion test, were not included in the CEL. Before including this additional area, TUEC identified on the CEL approximately 55,000 square feet of unqualified or inde-terminate coatings. This value is already considered high by the TRT, and it would be more than doubled by including the additional area. The TRT finds this value (about 20 percent of the total coated area at CPSES) to be excessive when compared to CEL areas reported by other applicants. Regarding the training and qualification of inspectors, the TRT finds a number of deficiencies. In many cases, records of inspector education, previous experi-ence, training, qualification testirg, and certification do not provide evidence adequate to demonstrate the capability of inspection personnel. The allegation that some instructor functions were performed by inadequately qualified per-sonnel was substantiated. The extent to which these deficiencies affected the quality of the completed coating work is indeterminate. 3.4 Overall Assessment and Conclusions The TRT evaluation of the protective coatings area revealed many specific defi-ciencies which render a relatively large percentage of the coatings at CPSES unqualified. However, consistent with the guidelines of the Standard Review Plan, Section 6.1.2, TUEC has provided justification that debris generated from the failure of all paint in the Containment Buildings under design basis acci-dent conditions will not adversely affect the performance of post-accident , fluid systems. In Appendix L to this supplement, the staff evaluates this justification and concurs with TUEC's conclusions. Therefore, a determination has been made that coatings inside of Containment Buildings do not need to be I qualified (see Appendix L). Comanche Peak SSER 9 M-10 l i
. . _ _ _ ._. . _ . . _ _ ~ . _ _ _ _ . . . . - _ _._.. ._ _
However, based on TUEC's prior FSAR commitment to provide qualified coatings, those applied before issuance of Appendix L were required to have been applied as qualified coatings. The failure of TUEC to fulfill that prior commitment indicates deficiencies in the coatings QA/QC program. The number and type of deficiencies found by the TRT evaluation clearly demon-strate serious weaknesses in the coatings QA/QC program in design analysis, material control, instructions, performance and inspection of the work, quali-fication of personnel, and documentation, all of which rendered the program
- inadequate to assure compliance with the requirements in effect at the time the work was performed. These deficiencies, although now determined not to be of safety significance in the coatings area, will be considered in evaluating the effectiveness of TUEC's overall QA/QC program.
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- 4. Actions Required of TUEC The actions required in the protective coatings area reflect the findings and conclusions of Appendix L to this supplement. The TRT does not recommend any actions to remedy deficiencies in the coatings applied at CPSES. However, actions are required to document the status of existing coatings so that future inspection and test programs implemented to comply with the guidelines of Appendix L to this supplement can be based on known coating conditions.
4.1 Backfit Test Program (See Coatings Category 1) Apply the Elcometer calibration correction to the data for the 4714 adhesion tests covering 2189 miscellaneous steel items tested to establish a more reliable estimate of the adhesion test failure rate. This revised analysis should include a statistical analysis showing the 95 percent con-fidence upper limit of the failure rate for all the miscellaneous steel items inside the Containment Building. Analyze the corrected data to establish a more reliable estimate of the fraction of tested miscellaneous steel coated surface that failed the adhesion test acceptance criterion. Enter the resulting failed areas in the protective coating exempt log. (See Coatings Category 6.) The coating exempt log will be used in planning future inspections of coatings consistent with the guioelines of Appendix L. 4.2 Traceability (See Coatings Category 3) CPSES Nonconformance Reports-(NCRs) C-81-01724 and C-81-01673 provide "use-as-is" dispositions for discrepant coating materials with inadequate technical justification for the disposition. Accordingly, provide adequate technical justification to demonstrate the acceptability of the batches of coating materials listed in these NCRs or, alternatively, identify.and quantify the areas where these batches were used and place these areas in the coatings exempt log. Additionally, review all other NCRs which report discrepant or irregular conditions in coating materials. For any such NCRs which were dispositioned "use-as-is," identify the batches and provide ade-quate technical justification for their acceptance,'or identify and quantify the areas where the batches were used and place these areas in the coatings exempt log. The coating exempt log will be used in planning future inspections of coatings consistent with the guidelines of Appendix L to this supplement. 4.3 Coatings Procedures (See Coatings Category 4) The TRT found deficiencies in procedures and instructions for coating work-and related inspection activities during the construction phase, which rendered them inappropriate or inadequate for determining satisfactory accomplishment of important activities. The TRT also found that the procedure review and approval system was inadequate to detect and correct these deficiencies. Comanche Peak SSER 9 M-13
Accordingly, make the necessary changes to the procedure review and approval system to assure review and approval by technically qualified individuals to prevent recurrence of the types of deficiencies discussed in Coatings Category 4, and to assure that procedures are reviewed for con-sistency and clarity. Apply this revised review and approval system to the , issuance and revision of all procedures which will govern future coating work, inspection, and testing at CPSES, consistent with the guidelines of Appendix L to NUREG-0797, Supplement 9. ' 4.4 Coatings Exempt Log (See Coatings Category 6) Provide updated estimates of the additional items including those detailed in Coatings Category 6 to be entered into the exempt log. Although all coatings are now considered exempt, maintain the CEL separately to identify all items which did not meet the requirements in effect at the time the coating work was performed. This log will be used in planning future in- . spection of coatings consistent with the guidelines of Appendix L of NUREG-0797, Supplement 9. t Comanche Peak SSER 9 M-14
ATTACHMENT 1 LIST Or TECHNICAL _ CONCERNS _AND ALLEGATIONS IN THE PROTECTIVE C0ATINGS AREA Allegation Page Number Characterization Category Number AQ0-1 Imperial coatings (Southern Imperial 2, 4 M-49, Coatings of New Orleans) applied in the M-75 sequential order #115/1201/11S/1201 or
#115/1201/11/1201, in accordance with
, B&R Procedure CCP-40, Paragraph 4.3.1.2, are not DBA qualified. AQO-2 Repair coating systems applied in sequences 2, 4 M-49, which are different from the original M-75 application sequences, as discussed in non-conformance report (NCR) #C83-01752, June 23, 1983, are not DBA qualified. AQG-3 Carboline Phenoline 305 (P-305) applied 2, 4 M-49, over another manufacturer's epoxy coating, M-75 in accordance with design change authoriza-tion (OCA) #17,142, Revision 2, is not DBA qualified. AQ0-4 Carbo 11ne Carbo Zinc 11 (CZ-11) topcoated 2, 4 M-49, with Imperial 1201, in accordance with M-75 DCA #12,374, Revision 1, is not DBA qualified.
- AQO-5 P-305 applied over Ameron Dimetcote 6 2, 4 M-49, i (0-6), in accordance with CPSES Procedure M-75 i
#CCP-30A, Revision 2, Paragraph 1.3.1, is not DBA qualified.
AQO-6 Imperial Nutec 115 surfacer applied over 2, 4 M-49, foreign objects embedded in concrete, in M-75 accordance with CPSES Procedure #CCP-40, Revision 5, Paragraph 4.1.1.3, is not DBA
- qualified.
AQ0-7 CPSES NCR #C83-01986 provides a repair 4, 5b M-75, disposition for cracking and flaking of M-107 ! i concrete coatings which will not remedy the cause of the deficiencies. t 1 Comanche Peak SSER 9 M-15 I l l l
ATTACHMENT 1 (Continued) Allegation Page Number Characterizatio_n C,ategory Number AQO-8 CPSES procedure #CCP-30, Revision 11, 4 M-75
, allows inorganic zinc primer to be applied ! over zine residue which will cause adhesion problems and prevent galvanic action.
AQ0-9 Inorganic zinc primer in some locations 2, 4 M-49, was applied in three coats contrary to M-75
- procedure QI-QP-11.4.5, Revision 27, and i is therefore not DBA qualified.
AQ0-10 Coatings applied to surfaces which were 2, 4 M-49,
. prepared by power tool cleaning were M-75 l' smoothed or polished and thus do not have adequate surface profile to assure adherence
] at DBA conditions. AQO-11 Primer applied to a thickness of 0.5 mils, 2, 4, Sc M-49, 4 in accordance with DCA #18,489, may be M-75, too. thin to be qualified for a DBA. M-111 AQO-12 Imperial coating system 11S/1201/11S/1201 2, 4 M-49, applied at a thickness of 102 mils, in M-75
! accordance with CPSES Procedure CCP-40, j Revision 5, Paragraph 4.3.1.2, may not be
, qualified for a DBA. AQ0-13 Coatings applied in the reactor core cavity 2 M-49 { which will be subjected to higher levels of neutron and gamma exposure than coatings in i other areas may fail during a DBA. I AQ0-14 Inspectors are prevented from writing NCRs Sa, 5b M-103, i and must instead write unsatisfactory irs. M-107 ] Once written, anyone can sign off on NCRs and irs. A past QC supervisor voided many NCRs. Due to a poor tracking system, irs
, can be lost.
AQ0-15 CZ-11 or Carboline 191 primer (191P) applied 2, 4 M-49, over P-305, and P-305 applied over Imperial M-75 )
- 1201, in accordance with CPSES Procedure CCP-30, Revision 11, Paragraph 4.4.3.0, i are not DBA qualified.
i i l l l
, Comanche Peak SSER 9 M-16 i
. - - - . _ _ - - - __ _,_e _, . - _ _ - , ,- _. . . , __ - . . _ . . _ _ - -
. ~ . . .
ATTACHMENT 1 (Continued) Allegation Page - Number Characterization Category Number AQ0-16 QC inspectors are being pressured and
- intimidated, which may result in coatings deficiencies.
AQ0-17 Tests of the cleanliness of compressed air 4 M-75 used for spray application of coatings were invalidated due to the practices of pro-duction personnel. AQ0-18 Inspectors were not allowed to identify 1, 4 M-23, visual defects during backfit inspections. M-75 AQO-19 Backfit inspection procedures are vague. 1, 4 M-23, M-75 AQ0-20 Adhesion testing of the protective coatings 1, 4 M-23, was not performed properly during the M-75 backfit test program. AQ0-21 Adhesion test data were not corrected for 1 M-23 calibration error. AQ0-22 Backfit program adhesion testing is per- 7 M-121 formed by coatings inspectors prior to completing training. AQ0-23 The coatings QC program at CPSES is 4 M-75 inferior to such programs at other nuclear power plant project because it does not permit use of the standard tests which have been used on other projects. AQ0-24 Coatings have been placed over rusty, scaly, 4, Sc M-75, unprepared metal surfaces inside pipe M-lll supports made of tube steel without end-caps and may come off during an accident. AQ0-25 A seal coat which should have been rejected Sa M-103 was improperly accepted by QC personnel prior to the finish coat being applied. AQ0-26 DCAs are not controlled. Sc M-111
- This allegation is associated with intimidation and has been transferred to the OI.
Comanche Peak SSER 9 M-17
ATTACHMENT 1 (Continued) Allegation Page Number Characte_rization Category Number AQ0-27 DCAs are originated and approved by Sc M-111 Engineering without QA/QC input. AQ0-28 DCAs are written to make conditions Sc M-111 . acceptable so NCRs will not be written. AQ0-29 DCAs are written rather than reworking Sc M-111 deficient areas to overcome problems that are identified by NCRs. DCAs are used to downgrade surface pre- Sc M-111 AQ0-30 paration and specification requirements from safety to nonsafety. AQ0-31 QC management told inspectors "not to worry" 4, Sc M-75, about difficult access areas, and to "do M- 111 the best you can." AQ0-32 Reading list contents were changed after 7 M-121 inspectors had signed the list. AQ0-33 A lead coatings inspector lacked the 7 M-121 qualifications to properly perferm his duties. AQ0-34 The requirements of American National .3, 4 M-65, Standards Institute (ANSI) Standard M-75 N45.2.2-1978 were not met for coating material storage. AQ0-35 Workmanship was poor because painters lacked 7 M-121 the qualifications necessary to produce quality work; and painter certification documentation was deficient. The traceability of coatings materials was 3, 4 M-65, AQ0-36 not always maintained. M-75 The backfit test program was improperly 1 M-23, AQ0-37 performed. Maps were incorrect and documentation was forged and falsified. The method used at CPSES to remedy high 4 M-75 AQ0-38 dry film thickness (DFT) of CZ-11 will burnish the primer and result in poor adhesion of the topcoat. Comanche Peak SSER 9 M-18
ATTACHMENT 1 (Continued) Allegation Page Number Characterization Category Number AQO-39 Applied P-305, one and two years old was 4 M-75 topcoated with new P-305 with little or no surface preparation. AQ0-40 Residues resulting from power tool cleaning 4 M-75 of surfaces were removed by improper methods which could leave contamination or debris under the top coat. AQ0-41 An improper cleaning solution which was used 4 M-75 to wipe surfaces immediately prior to repairs left behind prohibited impurities. AQ0-42 Imperial 115 and 1201 were applied over 4 M-75 duct tape and foam rubber on Richmond Inserts resulting in the appearance of a solid wall where, in fact, holes exist. AQO-43 The methods used at CPSES to verify the 4 M-75 cure of inorganic zinc primers are not i adequate, and inorganic zinc primers are not properly cured prior to topcoating. AQ0-44 The " nickel test" for verifying the cure 4 M-75 of inorganic zinc primers prior to top-coating was not performed properly. AQ0-45 Repairs of defects have been accomplished 4 M-75 with no reinspection of the repairs. AQ0-46 Some adhesion test samples showed unaccept- 1 M-23 , able substrate conditions, including rust. ' AQ0-47 QC inspectors were instructed to perform 1 M-23 approximately 25 Elcometer adhesion tests, in violation of written instructions. AQ0-48 Coatings were applied over seismic joints-4 M-75 which were filled with foam and were not to be coated. AQ0-49 Overspray was allowed and was commonplace 4, Sa M-75, in areas which had been inspected previously. M-103 AQ0-50 Coatings have been applied over steel Sa M-103 substrates without quality control inspection. Comanche Peak SSER 9 M-19
ATTACHMENT 1 (Continued) e , Page Allegation Category Number Number Characterizati_on Excessive' thinning of P-305 resulted in 4 M-75 AQ0-51
- a Weak and brittle film and made it impossible to obtain a Tooke gauge reading.
Coatings have been applied over concrete 5a M-103 AQ0-52 substrates without QC inspection. Sa M-103 AQ0-53 QC inspectors have been denied the oppor-tunity of writing Request for Information or Clarification (RFIC). M-23 AQ0-54 During the Backfit Test Program, only the 1 first unsatisfactory DFT reading was recorded even if subsequent readings were further ' outcof-specification; thus, adversely ' affecting the trend analysis. Areas identified during the Backfit Program 1 M-23 AQO-55 as requiring coatings removal did not have the coatings removed. AQ0-56 Original docurentation for the Sackfit , Program was-destroyed by QC management. An area at the 860-foot elevation in , Sa
, M-103 AQ0-57 Unit 2 had coatings applied over filth.
AQ0-58 CPSES QC inspection procedures require ' A M-75 that inspections be performed "at arm's length" and with inadequate. lighting. Substandard coatings on the liner plate 5a M-103 AQ0-59 were accepted by a QC.. inspector. AQ0-60 Coatings inspectors were selectively sent to various inspections so that areas would pass inspection. Prospective inspectors were sometimes 7 M-121 AQ0-61 trained by unqualified instructors and management had been aware of the practice.
*This allegation is associated with wrong-doing and has been transferred to the OI. .
Comanche Peak SSER 9 M-20 . m
ATTACHMENT 1 (Continued) Allegation Page Number Characterization Category _ Number AQ0-62 Some paint used at CPSES in Service Level I 3, 4 M-65, areas was contaminated with grease and oil M-75 prior to application and was applied anyway. Comanche Peak SSER 9 M-21
ATTACHMENT 2 ASSESSMENT OF INDIVIDUAL TECHNICAL CONCERNS AND ALLEGATIONS IN THE C0ATINGS AREA
- 1. Allegation Category: Coatings 1, Backfit Test Program
- 2. Allegation Number: Parts of AQ0-18, AQ0-19, AQ0-20, AQ0-21, AQ0-37, AQ0-46, AQ0-47, AQ0-54, AQ0-55 and AQ0-56.
- 3. Characterization: It is alleged that:
Visual defects were not identified during backfit inspections (AQ0-18). Backfit inspection procedures were vague (AQ0-19). Adhesion testing of the protective coatings was not performed properly (AQ0-20). I - Adhesion test data were not corrected for calibration error (AQ0-21). An area stated to have satisfactory documentation, in fact, had primer coatings exceeding the allowed thickness (AQO-37a). Maps for the backfit test program were incorrect (AQ0-37b). Documentation for the backfit test program was forged and falsified (AQ0-37c). . QC inspectors completed inspection reports (irs) without performing the inspections (AQ0-37d). Unacceptable substrate conditions were observed through Tooke gauge tests (AQ0-46). Adhesion test dollies used during the backfit test program were
- ' observed to have rust adhering to the paint (underside) at the com-pletion of the test (AQ0-46b).
~ Twenty-five adhesion tests were performed in violation of written instructions (AQ0-47). During the backfit test program, only the first unsatisfactory read-ing was recorded, which adversely affected the trend analysis (AQ0-54). Areas identified during the backfit test program as requiring coatings removal did not have the coatings removed (AQO-55). Original documentation for the backfit test program was destroyed by QC management (AQ0-56).
- 4. Assessment of Safety Significance: In order to assess the individual allegations characterized above, the NRC Technical Review Team (TRT) re-Comanche Peak SSER 9 M-23
viewed the background and scope of the backfit test program (BTP) and 4 independently evaluated the program test results as presented below. The TRT assessment of the individual allegations follows in Section 4.d.
- a. Backfit Test Program Background and Scope (1)
Background:
In 1981, Region IV (RIV) of the NRC inspected protective coatings at the Comanche Peak Steam Electric Station (CPSES). As a result of this inspection, which culminated in the issuance of Inspec-tion Report (IR) 81-15, RIV issued a Notice of Violation regarding the failure of Texas Utilities Electric Company (TVEC) to follow quality assurance program procedures for the inspection of protective coatings. Specifically, from late September 1979 through October 1981, docu-mentation for protective coatings inspections either was not main-tained or was incomplete. In response to the Notice of Violation, TUEC, in a letter to NRC dated November 19, 1981, proposed instituting a backfit test program (BTP) and documented the cited discrepancies as nonconforming condi-tions. TUEC also proposed a complete review of existing records and a reinspection (using destructive testing).of coated areas for which documentation was missing or discrepant. The reinspection was to be based on a statistically sound sampling plan. Both dry film thickness tests (Tooke gauge tests) and adhesion tests (Elcometer tests) were to be used to evaluate the condition of the applied coatings, and any discrepant areas were to be clearly identified and corrected in accordance with approved procedures. On January 19, 1982, the NRC responded to TUEC's November 19, 1981, letter. The NRC had no ques-tions at that time, and informed TUEC that they would review the corrective actions during a future inspection. In January 1984, RIV contractel with Brookhaven National Labora-tory (BNL) to provide techn~ical assistance for onsite reviews and technical evaluation of the allegations of deficiencies related to the protective coatings program at CPSES. As part of BNL's review work at CPSES, the NRC requested that BNL perform some independent testing of the protective coatings. These independent tests are discussed later in this report. On April 25, 1984, BNL sent an interim report on protective coatings to RIV. This report requested information from TUEC regarding the BTP, presented BNL's independent test results, and reported BNL'-s interim findings on TUEC's protective coatings procedures and docu-mentation. By memo dated May 22, 1984, from D. Eisenhut-to the ASLB, BNL's April 25, 1984, interim report was provided to the ASLB. On June 13, 1984, BNL sent a draf t " Status Report on Protective Coatings Allegations" to RIV, providing the current status of BNL's investiga- . tion of 60 allegations regarding protective coatings at CPSES. On i July 9,1984, the TRT assumed responsibility for the investigation of the protective coatings allegations, based on the E00 directive of March 12, 1984. By memo dated August 7, 1984, from D. Eisenhut to the ASLB, the status of the 60 allegations under investigation by BNL were provided to the ASLB. These 60 allegations plus 2 additional alleg&tions formed the basis for the TRT coatings investigation. In Comanche Peak SSER 9 M-24
BNL's April 25, 1984, interim report of their investigation, deficiencies were noted in " Testing, Procedures, Documentation / Design Control." The BNL interim report concluded that: "The coatings procedures and design control for coatings at CPSES appear to be inadequate to assure the specification of proper coatings systems and the application of coatings, once they are specified." In addition to the review of individual allegations, the TRT conducted an independent generic review of coatings, as discussed in Section 3.1 of this supplement. The TRT overall assessment and conclusions, based on its generic review (Section 3.4 of this Appendix), states: The number and type of deficiencies found by the TRT evalua-tion clearly demonstrate serious weaknesses in the coatings QA/ QC program in design analysis, material control, instructions, performance and inspection of the work, qualification of per-sonnel, and documentation; all of which rendered the program inadequate to assure compliance with the requirements in effect at the time the work was performed. Therefore, the TRT's conclusions are in substantial agreement with BNL's interim assessment, even though the TRT did not compare each allegation it assessed with those assessed in BNL's interim report. The TRT substantiated or partially substantiated most allegations, including those examined by BNL. However, the TRT found that none of the allegations except the one related to the backfit test program were of significant technical concern, either because they were not substantiated or because they involved relatively small areas of coating. The one exception pertains to coatings on mis-cellaneous steel items which failed the adhesion test. These areas may amount to 6 percent of the total area coated. (2) Scope: The purpose of the BTP was to review and reinspect the coated steel liner, concrete, and miscellaneous steel for which coatings documentation was missing or discrepant. Tooke gauge tests were used to measure the dry film thickness (DFT) of the primer and topcoat, and Elcometer adhesion test results were used to determine if protec-tive coatings adequately adhered to the substrate. TUEC also selected a 200 psi pull criterion for the adhesion test, presented in Section 6,
" Physical Properties Tests," of ANSI Standard N5.12-1974, " Protective Coatings (paints) for the Nuclear Industry." TUEC intended to use the DFT and adhesion test data in lieu of the missing and discrepant documentation.
It is the TRT's opinion that if DBA qualified paint, as defined in the introduction to Appendix L, was applied and its traceability was maintained, the BTP, properly administered, could provide useful, indirect information on the quality and design basis accident (DBA) survivability of the coating work with missing or discrepant docu-mentation. The DFT tests could demonstrate that the protective coatings were applied in the same thickness ranges (for primer and topcoat) as the DBA qualification tested samples. Adhesion tests Comanche Peak SSER 9 M-25 i t _ . - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - - - _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - ._ . _ _ _ _ _ __-_- _..___ -_____ _ _-_- i
provide data which indirectly demonstrate the adequacy of surface preparation, in that the primary purpose of surface preparation is to t provide good coatings adhesion. The adhesion tests also provide some assurance that application and curing of the coatings are adequate to produce satisfactory coating film integrity and internal strength. The adhesion criterion of 200 psi is appropriate, because this criterion applies to physical properties testing of DBA qualification test samples in accordance with ANSI N5.12.
- b. Evaluation of Backfit Test Program (BTP) Results (1) General: In its review of BTP inspections and tests, the TRT paid particular attention to the adhesion test results because they pro- ,
vided the most direct indication of coating adherence to Containment Building surfaces under accident conditions, the primary safety con-cern related to coatings. Adhesion testing was performed using an Elcometer adhesion tester,- which measures the force required to pull a protective coating off the coated surface. The Elcometer model used by TUEC inspectors for the BTP had a total range of 0 to 1,000 psi, and could be read in the field with a precision of approximately 50 psi. The Elcometer ' readings tended to read high after repeated use; therefore, the Elco-meters were periodically recalibrated by deadweight testing in the onsite Brown & Root instrument shop. 3 The partial results of the adhesion tests and dry film thickness tests on the Containment Building steel liner, the concrete surfaces, and the surfaces of miscellaneous steel components in Unit 1 were re-ported by correspondence from L. Bielfeldt of TUEC to D. Lurie and L. R. Abramson of NRC, dated March 29, April 17, and April 23, 1984. The failure rates (Elcometer readings below 200 psi) were very low. For the Containment Building liner, only 2 out of 405 paint samples failed; for the concrete surfaces, there were no failures in 1,691 readings; and for the miscellaneous steel surfaces, there were 20 failures in 1,517 readings. As discussed in detail below, the DFT test failures rates were also low. On the basis of these low failure rates, TUEC, in a memorandum from R. G. Tolson to its inspection staff, on February 10, 1984, discontinued all-routine destructive testing (adhesion tests and Tooke gauge coating thickness tests). In a letter to the NRC dated February 15, 1984, TUEC referred to verbal notification on January 16, 1984, to Mr. R. G. Taylor of the NRC, of a " deficiency regarding an error in the tolerances used in the calibration of the adhesion tester." At the July 11, 1984, site meeting, TUEC briefed the TRT on the overall scope of the coating backfit test program. R. Tolson (TUEC) informed the team of the discrepancy in calibrating Elcometers used for the coating adhesion
~ test. This discrepancy, which was discovered after most of the BTP adhesion tests were completed, would allow in plant test results to be too high by as much as 200 psi. Thus, any Elcometer reading less than 400 psi represented a potentially failed area of coatings.
Comanche Peak SSER 9 'M-26
- - - _ _ _ _ _ _ _ _ _ _ _ _ _ - ______________.__1___ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
After learning of th;s deficiency, the NRC, by attachment to an NRC meeting notice memorandum dated July 27, 1984, requested that TUEC provide the TRT with corrected adhesion test data and analyses of failure rates for the containment liner, concrete, and miscella-neous steel. In the TRT's opinion, the corrections to the Elcometer readings could reasonably be made from calibration data available in the instrument shop for each Elcometer for each date on which it was checked. Part of this information was provided by TUEC in a letter to NRC Region IV, dated July 16, 1984, which was prepared in response to a previous NRC request of May 23, 1984. TUEC's correspondence con-tained the calibration data for each Elcometer and the majority of the protective coatings inspection reports (PCRs) for the backfit test program. (Although.TVEC sub.mitted 278 PCRs, its liner map indi-cated that there were 339 PCRs.) TUEC's transmittal letter stated that the package contained results for 869 adhesive tests for the liner, 2,128 tests for the concrete, and 4,714 tests for the miscel-laneous steel. The PCRs recorded the original, uncorrected adhesion data and the dry film thickness data. A list of PCRs for the Containment Building steel liner coatings with adhesion test readings below 200 psi after correction, and an analysis of the failure rate, were transmitted to NRC by TUEC in a letter dated August 14, 1984; the letter included an interoffice memorandum from R. C. Levine to R. G. Tolson, dated August 10, 1984. For the concrete surfaces in the Containment Building, another TUEC inter-office memorandum, from R. G. Tolson to file, dated September 10, 1984, was submitted to the TRT, and provided a list of PCRs with adhesion test readings below 200 psi after correction, and an estimate of failure rate. (2) Liner Plate Test Results (a) Adhesion Testing. TUEC found that 51 out of 869 adhesion test readings, or 5.9 percent, for the coatings on the Containment Building liner were below 200 psi after correction for calibra-tion error. The failed area, calculated by summing the areas corresponding to each failed reading, was 5,148 square feet, or 3.5 percent of the total liner surface, which TUEC estimated to be 145,088 square feet. In this evaluation, the TRT expresses failure rates in terms of area failed rather than in terms of number of failed tests because each test may represent a differ-ent number of square feet. Test areas typically range from approximately 20 square feet to 145 square feet. The TRT believes that TUEC's estimate of total liner surface is too high. Considering the liner surface as a cylinder (diameter 135 feet, height 192 feet) topped by a hemisphere (diameter 135 feet), the surface area is 110,000 square feet. TUEC per-sonnel interviewed by the TRT were unable to explain why they calculated the area to be 145,088 square feet. TUEC also estimated that 96 percent of the total liner surface was backfit tested. Based upon its review of the liner surface Comanche Peak SSER 9 M-27
backfit test map, the TRT found the 96 percent estimate to be 4 reasonable, so that the liner surface tested for adhesion would be 106,000 square feet. Using this figure, 4.8 percent of the area tested failed the adhesion test. l The TRT independently assessed the liner failure rate from the
- original adhesion data after correcting.the readings according to the Elcometer calibration data from the instrument shop.- The 4
TRT found that, after correction, 36 adhesion test readings (out of a total of 834) were below 200 psi, giving a test failure
- rate of 4.3 percent. The failed area, calculated by summing the areas corresponding to each failed reading, totalled 3,092. square feet or 2.9 percent of the area tested.
i The package of PCRs delivered by TUEC contained only 278 PCRs, with 3 adhesion readings recorded on each PCR, for a total of 834 adhesion test readings. The TUEC transmittal letter of August 14, 1984,- for the package and the memorandum of August 10, 1984, referred to 869 adhesion test readings, which corresponded to about 290 PCRs, assuming that the usual 3 readings were reported on each PCR. However, 339 areas on the Containment Building liner map provided by TUEC were labeled with different-
- PCR numbers. Apparently, not all of the liner PCRs were included in the package delivered to the TRT, nor were they included in the group discussed in the August 10, 1984, TUEC memorandum.
] TUEC personnel. interviewed by the TRT were unable to account for the discrepancy, and agreed that the 339 areas on the liner map represented the. total number of PCRs. On the assumption that the average liner area per PCR was the - l same for the missing PCRs as it was for those delivered to the TRT, the adhesion-tested area would be 106,000 square feet multiplied by 278/339, or 87,000 square feet. Using this area, the TRT found that 3.6 percent of the liner area tested failed the adhesion test. The TRT assumed the same failure rate for the liner area represented by the missing PCRs. When a similar l correction is applied to the area failure rate calculated by TUEC, the TUEC failure rate becomes 5.9 percent of the tested area. A more accurate estimate of-the area represented by the 4 missing PCRs can be obtained by locating.the missing PCR areas on the liner map and summing them. Without the missing PCRs themselves, an accurate estimate of adhesion test failure rate for these areas cannot be obtained. To account for the discrepancy between the number of failures found by the TRT, 36 or 3,6 percent, and by TUEC, 51 or 5.9 per-cent, the TRT examined the methods of correcting the original adhesion data for calibration error. The TRT determined that TUEC was more conservative (used larger corrections) in correct-
- ing for calibration errors, so that some adhesion test readings which were below 200 psi after the TUEC correction were above 200 psi after the TRT correction.
(b) Dry Film Thickness (DFT) Testing. A-second concern evaluated-in determining the quality of liner. coatings for the BTP was the Comanche Peak SSER 9 M-28 l l 1
r dry film thickness (DFT) of the primer coat and of the total l i coating system. These thicknesses are measured with a Tooke l gauge. With this device, a sharp V-shaped cut through the coatings to the substrate is made, after which the edges of the cut are examined with the optics of the instrument to determine the minimum, maximum, and average thicknesses of the primer and of the total coating system. The original TUGC0 backfit inspection procedure for protective coatings on steel, QI-QP-11.4-23, issued on November 19, 1981,
- specified the following acceptable thickness ranges
Single Average of 5 Coating Reading, Mils Readings, Mils Primer - Carboline CZ-11 1.5 - 5.5 2.0 - 4.5
"- Ameron 06 1.5 - 5.5 2.0 - 5.0 Total System -
Primer & Phenoline 305 topcoat 7.0 -11.5 7.0 -11.0 A majority of the liner PCRs reported at least one DFT value outside the original specifications. Subsequent revisions of the procedures broadened the acceptable thickness ranges. The latest version (Revision 13, dated April 18, 1984) listed the following ranges: Single Average of 5 t Coating Reading, Mils Readings, Mils Primer - Carboline CZ-11 1.5 - 8.0 1.5 - 7.0
- Ameron 06 1.5 - 7.0 1.5 - 7.0
- Carboline 191 1.5 - 7.0 1.5 - 7.0 Total System . From full " hiding" From full " hiding" Primer & Phenoline by topcoat to less by topcoat to less 305 topcoat than 15.0 than 13.0
' (TRT technical concerns involved in enlarging the acceptable range of thicknesses, as related to DBA qualifications of coat-ings, are discussed in Coatings Category 2 of this Appendix.)
l Even after the DFT specifications were broadened, many of the liner PCRs contained one or more DFTs outside the allowable range, most often on the low side for the total system (primer plus topcoat). As a consequence, according to interviews with , TUEC QC personnel, nearly all of the liner coatings in place at the start of the backfit test program were reworked. The TRT review of records of the disposition of NCRs confirmed rework for many liner areas. The repairs were performed in accordance with the approved Texas Utilities Generating Company (TUGCO) repair procedure (QI-QP-11.4-23) and were reinspected. 1 A few small, unrepaired liner areas (totalling about 110 square , feet) were placed in the coatings exempt log (CEL) as Items 8 j Comanche Peak SSER 9 M-29
to 18. (The TRT review of the CEL is discussed in Coatings l Category 6 of this Appendix.) When the repairs were completed, less than one percent of the Containment Building liner coatings did not meet the DFT speci-fications of TUGC0 procedure QI-QP-ll.4-23, Revision 13. (3) Concrete Test Results , (a) Adhesion Testing. As stated in 4.b(1), TUEC provided the results of 2,128 adhesion tests on concrete coatings (634 PCRs, with 1 to 5 adhesion tests recorded on each) in their letter dated July 16, 1984. TUEC stated that the 2,128 tests represented approximately 50 percent of the concrete surfaces in the Containment Building. The concrete surfaces not subjected to backfit inspection were: (1) areas not coated at the time of backfit inspection; (2) areas inaccessible to test equipment; and (3) areas not inspected (due to termination of the test program on February 10, 1984). As discussed in 4.b(1), TUEC's initial reports on the adhesion tests on concrete indicated no failures in 1,691 tests. When the Elcometer readings recorded on the PCRs were corrected for calibration error, TUEC (in a memorandum from R. G. Tolson to File, dated September 10, 1984) reported 65 adhesion test readings out of 2,128 tests with values below 200 psi, which corresponds to a failure rate of 3.1 percent. TUEC provided l neither a calculation of total concrete area with coatings failing the adhesion test nor a calculation of the percentage of concrete area where coatings failed the adhesion test. The concrete surfaces, because of their more complicated geometry, are less amenable to accurate determination of the tested sur-face area than the Containment Building liner surfaces. Since
~
test procedure QI-QP-11.4-24 called for approximately one adhesion test and one set of 0FT readings per 100 square feet of concrete surface, the failure rate in terms of area can be approximated by the test failure rate. The TRT did not conduct a complete independent analysis of the massive amount of adhesion data for the concrete coatings. In-stead, the TRT elected to restrict its audit of the adhesion data to the coatings on the interior surfaces of the concrete compart-ments surrounding steam generators No. 1 and No. 4. These sur-faces were selected, in part, because of their proximity to the Unit 1 Ccntainment Building sump screens. However, an analysis subsequently performed for Appendix L of this supplement demon-strated that this debris would not reach the sump screens. Areas of the interior walls of steam generator compartments No. 1 and No. 4, which had been backfit-inspected, were de-lineated and labeled with PCR numbers on TUGC0 drawings of these surfaces (drawings PCRM-018A1 and PCRM-01881 for compartment 1; drawings PCRM-019Al and PCRM-01981 for compartment 4). From the data recorded on each of these PCRs (20 in compartment 1 and 11 in compartment 4), the TRT tabulated the date of testing, the Comanche Peak SSER 9 M-30
Elcometer readings, the Elcometer used, and the area tested. For the concrete adhesion tests, the number of adhesion tests per PCR varied from 2 to 5, with approximately one test per 100 square feet of sampled area. The TRT corrected the original adhesion test results for calibration error from the calibration data provided by the instrument shop. After correction, the TRT found 8 adhesion tests reading 200 psi or lower out of a total of 116 tests, giving a test failure rate of 6.8 percent. Three of the eight failures were within 10 psi of 200 psi; if these were not counted as failures, the test failure' rate would be 4.3 percent, which is in better agreement with TUEC's estimate of 3.1 percent. The total failed area, calculated by summing the areas corre-sponding to the failed tests, was 745 square feet. The TRT assumed in calculating this area and the total tested area that each recorded set of DFT tests corresponded to 100 square feet. The total tested area represented by the 31 PCRs was approx-imately 11,000 square feet. The TRT, therefore, found that 6.8 percent of the area of protective coatings on the interior concrete surfaces of steam generator compartments Nos. 1 and 4 failed the adhesion test. If the three borderline failures were neglected, the failure rate by area would be 4.0 percent. Based on TUEC's estimate of 285,000 square feet for total concrete surface area, the TRT's 6.8 percent failure rate corresponded to a failed concrete coated area of 19,400 square feet, whereas TUEC's 3.1 percent failure rate corresponded to 8,800 square feet. The TUEC letter of July 16, 1984, stated that approximately 50 percent of the concrete area was backfit inspected. The TRT estimated from the drawings of steam generator compartments Nos. 1 and 4 that the total area of the internal surfaces was approximately 20,200 square feet. .The backfit-tested area was approximately 11,000 square feet, or approximately 55 percent of the total area, which was consistent with the TUEC estimate . that 50 percent of the total concrete area in containment was backfitted. (b) Ory Film Thickness (DFT) Testing. The original TUGC0 backfit inspection procedure for protective coatings on concrete, QI-QP-11.4-24, Revision 0 (February 5, 1982), specified the acceptable coating thickness range for Reactic 1201 topcoat on concrete as a minimum of 3 mils and a maximum of 12 mils. As required by procedure; five scratches, spaced randomly over each 100 square feet of the sampled concrete area, were made with the Tooke DFT tester. A single reading was selected as representa-tive of coating thickness of each scratch. The " minimum" re-corded on the PCR was the lowest of the five readings; the " maxi-mum" was the highest; and the " average" recorded was the average of the five. The permissible " maximum" thickness limit was ex-panded to.16 mils in Revision 3 of the procedure on June 29, 1982. l
- Comanche Peak SSER 9 M-31
In correspondence dated April 17, 1984, from L. Bielfeldt of TUEC to D. Lurie and L. R. Abramson of NRC, TUEC reported that 101 recorded DFT readings out of a total of 4,623 on concrete coatings failed to meet the thickness specifications given above, giving a test failure rate of 2.2 percent. Most of these failures were for low topcoat thickness. In a manner similar to that for the adhesion tests on concrete, the TRT restricted its audit of the backfit DFT tests to the coatings on the interior concrete surfaces of steam generator compartments Nos. 1 and 4. The TRT examined the DFT data recorded on the 31 PCRs for these surfaces and found 10 recorded DFT readings out of a total of 297 which failed, giving a fail-ure rate of 3.3 percent. Only'one of the failed tests exceeded the allowable thickness; the remaining failures had thicknesses below the allowable 3 mils minimum. It is not clear why the DFT failure rate in steam generator compartments Nos. 1 and 4 was greater than the overall DFT failure rate reported by TUEC. The failure rate may be high because the small area sampled was not representative of the total concrete area. In any case, according to statements by
- TUEC QC personnel and the TRT inspection of the disposition of NCRs for many concrete areas, nearly all of the concrete areas with failed DFTs were repaired according to TUGC0 procedure QF-QP-11.4-24 until the DFTs were satisfactory. The TRT conducted i a random sampling of irs and travelers which confirmed this.
! The principal exceptions were the coatings on the concrete j surfaces of the reactor cavity (3,135 square feet) and the coatings on the interior of the elevator enclosure (2,700 square feet). These areas were placed in the CEL. The bulk of the concrete coatings in the Containment Building either had satis-factory DFTs on the first backfit inspection or were repaired until they passed the DFT test. (4) Miscellaneous Steel Test Results (a) Adhesion Testing. As stated in 4.b(1), TUEC provided RIV with , the results of 4,714 adhesion tests on miscellaneous steel coat-ings (2,189 PCRs recording 1 to 3 adhesion tests on each) in a package transmitted by letter, dated July 16, 1984. TUEC stated that the 4,714 adhesion tests represented approximately 22 percent of the coated miscellaneous steel surfaces in the Containment Building. The miscellaneous steel category includes items such as pipe supports, cable tray supports, and conduit supports. The surfaces not subjected to the backfit inspection were: (1) those not coated at the time of backfit inspection; (2) those inaccessible to test equipment; and, (3) those not i inspected due to termination of.the test program on February 10, 1984. The TUEC letter of July 16, 1984, indicated 26 failures out of the 4,714 adhesion test readings, giving a test failure rate of 0.55 percent. TUEC did not provide an analysis of the adhesion Comanche Peak SSER 9 M-32
test data on miscellaneous steel after correcting for Elcometer calibration error. The TRT did not conduct a complete independent analysis of the massive amount of adhesion data for miscellaneous steel. The data recorded on a randomly selected group of 42 PCRs, encompas-sing 78 adhesion tests, were analyzed in detail. The surface areas of 22 of these items were recorded as less than 10 square l feet, on which only one adhesion test was usually made. The TRT observed that, after correction for calibration error, 15 ; adhesion test readings were below 200 psi, giving a test failure l rate of approximately 19 percent for the 78 adhesion tests. Be- l cause of the large variation and uncertainty in the area repre- l sented by each adhesion test for miscellaneous steel items, the l TRT did not attempt to determine a failure rate in terms of area from the available data. The TRT failure rate was so much larger than the rate reported by TUEC (0.55%) due to the effect of correcting for calibration error. Only 1 of the 15 failed readings observed by the TRT was less than 200 psi before correction for calibration error, re-sulting in an uncorrected failure rate in fair agreement with TUEC's uncorrected rate. The 42 PCRs chosen by the TRT for detailed audit represented a. small fraction of the total of 2,189 PCRs on miscellaneous steel'. It is, therefore, quite possible that the selected group was not representative of the total population of miscellaneous steel items. However, the TRT's uncorrected failure rate did approxi-mately agree with TUEC's uncorrected rate for the total popula-tion. In the July 16, 1984, package of backfit data provided by TUEC, the total area of miscellaneous steel was given as 180,080 square feet. On the basis of the corrected 19 percent failure rate obtained by the TRT, the failed area of miscellan-eous steel could be approximately 35,000 square feet, assuming the TRT sample is a representative sample. (b) Ory Film Thickness (DFT) Testing. A TUEC analysis of the DFT results on miscellaneous steel was reported in correspondence, dated April 17, 1984, from L. Bielfeldt, TUEC to L. R. Abramson, NRC. The allowable thicknesses for coatings on steel are given in 4.b.(2), above. The TUEC analysis indicated that out of a total of 1,517 readings, 129 0FTs were outside the acceptable range, giving a test failure rate of 8.5 percent. , The TRT did not attempt to conduct a complete independent analy-sis of the massive amount of DFT data for the miscellaneous steel category. The set of 42 PCRs examined for coating adherence was selected to audit the DFT data as well. The TRT observed that out of a total of 252 readings, 39 DFTs were outside the allow-able range, giving a failure rate of 15.5 percent. Nine of the failed readings exceeded the allowable DFT; the remaining 30 were too low. Comanche Peak SSER 9 M-33
. - - - - - ~ - _ .- - _ - - . . - - _ -
d
; According to TRT interviews with TUEC quality assurance person-nel, most of the miscellaneous steel surfaces were reworked with additional topcoat, partly' for cosmetic reasons, so that the final DFT failure rate was lower than reported above. The.TRT conducted a random sampling of irs and travelers which confirmed this.
(c) Brookhaven National Laboratory (BNL) _ Adhesion Test Results. As discussed in section 4a, BNL and its consultant, under contract to Region IV of the NRC, performed some adhesion tests on randomly selected areas of the Containment Building liner, the i concrete surfaces, and the miscellaneous steel surfaces. Only ten adhesion tests were made on each of the three surface types, although it was recognized that such a small sample would have limited statistical significance. As reported in BNL's interim report to NRC Region IV, dated April 25, 1984, four out of the ten adhesion tests for the liner ! plate failed the 200 psi acceptance value with corrected read-4 ings of 156, 186, 186 and 186 psi. No failures were observed in the ten tests for the miscellaneous steel coatings. One of the ten tests failed for the concrete surfaces; however, this failure j was in the concrete substrate, not in the protective coating. Comparing these results with those reported by TUEC and with the i TRT audit of TUEC's results, the BNL results are consistent with i the corrected TUEC data for the concrete surfaces and for the miscellaneous steel coatings, considering the small size of the BNL sample. However, the 40 percent failure rate on the Contain-
! ment Building liner coatings observed by BNL was much higher than the failure rate of approximately 5 percent based on cor-rected TUEC adhesion test data.
-In a letter dated July 20, 1984, TUEC stated that three of the four failed liner tests sampled an area under the equipment :
hatch (elevation 812 feet, azimuth 225 ) which had not been ' painted until after the backfit program had been terminated. The TUEC letter stated that further adhesion testing showed that the failed area was extremely isolated and had been repaired. The TRT examined inspection records, including maps, pertaining l to the additional adhesion testing and repair in the liner area
- under the equipment hatch (PCR1-0031601 and. traveler VI-006720),
i and confirmed that the area failing the adhesion test was limited to a few square feet. In interviews with TUEC QA personne4, the TRT learned that the equipment hatch area was one of three small i liner areas in Unit 1 that was coated after the backfit program was terminated. At this time, only a nickel test was used to determine when the primer coat was sufficiently cured to permit i topcoating. (See Coatings Category 4, AQO-44, for a discussion of nickel test problems.) Also, during the BNL adhesion testing, BNL and TUEC QA personnel detected a solvent odor on the test dollies after they were pulled off the coated surface. This
- suggested a primer curing problem, which may have accounted for the high failure rate in the equipment hatch area.
! Comanche Peak SSER 9 M-34
Only one of the five adhesion tests of the liner at the 945-foot elevation failed. The TRT found from inspection records and a map of the additional TUEC adhesion testing (PCR1-0031602) that the failed area comprised only a few square feet and that adhe-sion tests within a foot or two on all sides gave readings above 200 psi. The TRT concluded that the high adhesion test failure rate ob-served by BNL for the containment liner represented a very limited area which was probably topcoated before the primer coat had cured. Therefore, the liner failure rate of approximately 5 percent, based on the corrected TUEC data on 869 adhesion tests, was more representative of the condition of the liner coatings as a whole. (d) Assessment of Indiv_i_ dual _ Allegati_ons. The TRT investigated specific allegations related directly or indirectly to the backfit test program. The TRT assessment of these allegations is discussed in the following paragraphs. (1) Visual Defects (AQO-18). It is alleged that QC inspectors are not allowed to identify visual defects, such as cracking or blistering, during backfit inspections. TUEC stated in its June 22, 1984 response to the NRC that "It was not intended that visual inspections of coated surfaces be performed as part of the backfit inspection program," which is incorrect. NCR C-81-01567, Rev.1 (dated November 22, 1981), NCR C-81-01373, Rev. 2 (dated November 3,1981), and NCR C-81-01613, Rev.1 (dated February 5, 1982), are generic NCRs covering the BTP, and require visual inspections as part of the NCR disposition. Instruction QI-QP-11.4-23, Rev. 2 (dated December 17, 1981), required a visual inspection; Rev. 3, deletes this require-ment. Similarly, Instruction QI-QA-11.4-24, Rev. 0 (dated February 5, 1982), required a visual inspection which was deleted in Rev. 1. The TUEC June 22, 1984, response also stated that visual inspections are part'of the finish-coat final acceptance inspections in procedures QI-QP-11.4-5 and QI-QP-11.4-10. The TRT reviewed procedure QI-QP-11.4-5, Rev. 5 (dated November 18, 1981) through Rev. 27 (dated November 18, 1983) and Procedure QI-QP-11.4-10, Rev. 2 (dated November 11, 1981) through Rev. 18 (dated January 16, 1984). Both of these procedures have steps for visual inspections included in the inspection report (IR). The TRT found that, although visual defects were not recorded as part of the backfit inspection, they were recorded using irs and other procedures. TRT concerns about procedures QI-QP-11.4-5 and QI-QP-11.4-10 and visual inspections are discussed further in the TRT's procedures review. (See Coatings Category 4 in this Appendix.) Comanche Peak SSER 9 M-35
(2) Backfit Pr: gram Vague __(AQO-19). It is alleged that In-structions QI-QP-11.4-23 and QI-QP-11.4-24 are very vague regarding the way the backfit inspections are to be conducted. A detailed evaluation of these procedures is found in Coatings Category 4 under the discussion of AQ0-19, which is summarized below. The TRT reviewed both QI-QP-11.4-23 (through Rev. 12, dated October 24, 1983) and QI-QP-11.4-24 (through Rev. 6, dated July 14, 1983). It was the TRT's opinion that these pro-cedures could be interpreted properly by a well trained QC Inspector, but that a QC Inspector who was not well trained in these procedures could have difficulty implementing them correctly. The TRT had additional concerns about TUEC coat-ing procedures and inspector training. (These concerns are discussed in Coatings Categories 4 and 7.) (3) Improper Adhesion __ Testing _(AQ0-20). It is alleged that ad-hesion testing of the protective coatings is not performed properly. It is also alleged that QC Inspectors are instructed not to cut around the adhesion test dollies when conducting adhesion tests contrary to the instructions pro-vided by the manufacturer of the adhesion tester. (The manufacturer's instructions are referenced by CPSES Speci-fication AS-31.) In the June 22, 1984, response to the NRC, TUEC stated that they did not follow the manufacturer's instructions that came with the instrument. The TRT determined that the allegation, as characterized above, was substantially correct. QC Inspectors did not scribe around dollies, although the manufacturer instructs the user to score around the dollies prior to performing the tests. TUEC provided results of onsite testirg to support the position that scribing or not scribing does not affect test results. The TRT reviewed these test results, as well as information from other qualified individuals in the industry, and determined that the failure to scribe around dollies did not affect test results and was not a technically improper procedure. The TRT noted that TUEC his since modified site practice to institute scoring around dollies in compliance with the manufacturer's instructions. (This allegation is also discussed under Coatings Category 4.) (4) Adhesion Tester Calibration Correction (AQO-21). It is alleged that Brown & Root is doing the calibration on these adhesion testers and that they are not using a corrected value curve (which should have been supplied with each unit) after each calibration. As alleged, TUEC did not correct adhesion test readings for 2 Elcometer calibration error until after the backfit in-spection program was terminated. The effect of making these corrections on adhesion test failure rates is dis- , cussed in detail in 4.b(1), above. Comanche Peak SSER 9 M-36 ! l k _ _ _ _ _ _ _ - _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
(5) Backfit Records (AQ0-37). It is alleged that: (a) in the backfit test program, areas that were docu-mented as having satisfactory primer actually had 10 mils of primer, which exceeded the allowed maximum; (b) none of the maps documenting areas of adequate primer are correct; (c) documentation for the backfit test program was forged and falsified; (d) a QC inspector on the night shift wrote up acceptable irs for the Containment Building dome area without ever performing those inspections; several QC inspectors would " buy off" anything; and, en several occasions at least one QC inspector conducted his coatings in-spections from several floors below where the paint was being applied.
" Satisfactory" primer exceeded the allowed maximum. The TRT discussed this issue with TUEC management, and with Quality Engineering and Quality Control inspectors. All parties contacted indicated that it was true that, in some cases, during backfit inspections performed on steel liner, spot DFT readings for primer coated surfaces with acceptable documentation were found to be outside of the acceptable DFT range of the primer and were documented as unsatis-factory. In its June 22, 1984, letter to NRC, TUEC stated that unsatisfactory readings detected during the backfit test program were tracked by irs unti? the conditions were corrected. The examination of backfit IR records indicated that unsatisfactory conditions were noted and a new IR was generated after the item was corrected. A random sampling of irs by the TRT confirmed this. The TRT determined that an alternate disposition was to prepare a design change authorization (DCA) that widened the range for the DFT readings. (TRT concerns with these DCAs and their effect on DBA qualification of coatings are discussed in Coating Category 2.)
Allegations AQ0-54 and -55, which are closely related to this subject are assessed individually below. Backfit liner maps incorrect. The TRT initially examined the liner plate maps used to keep track of backfit testing and rework. These maps indicated areas that had primer applied, inaccessible areas, and uncoated liner plate. These areas were clearly indicated by bold, dark border lines with the backfit inspection report (PCR) numbers, azimuths, and elevations identified in the center. There l were no apparent discrepancies noted when reviewing the l maps and associated PCRs. The only discrepancy that could { have occurred was that incorrect azimuths were given by the i Comanche Peak SSER 9 M-37
L QC inspectors. The TRT discussed this issue with QC j inspectors who were familiar with the mapping system. The j QC inspectors stated that there were no azimuth markings on the liner which QC could use to obtain precise measurements for location; therefore, QC inspectors used approximate , 4 locations. This practice would cause a gap or overlap t i between adjacent inspected areas when laid out on maps, ; even though there was full coverage of inspection performed on the liner by QC. i ? The TRT reviewed the final liner plate map which outlined ! all liner areas that were backfit tested. As stated in - 4.b(2) above, the TRT reviewed the data in 869 liner plate PCRs, compared their indicated locations with the final map, j and did not find any serious discrepancies. This map showed } that 96 percent of the liner plate area had been backfit t f tested. j Backfit test program documentation was forged or falsified. This allegation was investigated by the NRC Office of In- ! l vestigation (01) for possible wrongdoing. i The TRT reviewed the technical adequacy of the coating f j records and observed that there were photocopies of
- i paint-batch mix sheets attached to separate coating appli-cation inspection checklists for different items which
! appeared to be coated at the same time and in the same I environment. In each exhibit, there was some common identifier, such as Containment Building liner location or item description, that appeared to connect the two documents.
The use of photocopied paint-mix sheets was not a violation of procedure, and the TRT did not observe any technical inconsistencies in the exhibit inspection reports. The TRT l followed up the review of these exihibits by reviewing l three additional folders which contained approximately 250
- inspection reports that were randomly selected from the j site QA record vaults. Of this sample, the TRT found two inspection reports for which the coating mix date and application date were inconsistent. These two reports applied to the Containment Building liner and referred to i
NCR C-81-01567, which dispositions such items to be backfit inspected. Based on its review, the TRT could draw no conclusion as to falsification or forging of records. In addition, the two instances of technical inconsistencies which were found
; applied to the Containment Building liner, 96 percent of
, which was backfit inspected. Consequently, the TRT believes
' that the coated area associated with technically inconsis-tent records is very small.
A QC Inspector wrote up acceptable irs without physically 1 j inspecting the work. Also, several QC inspectors would j
" buy off" anything. The IRT group leader contacted the Comanche Peak SSER 9 M-38 l
alleger (A-31) by telephone on September 10, 1984, to obtain further details to aid in investigating this allega-tion. The alleger reiterated the names of the two inspec-tors that he had provided to NRC Region IV OI in October 1983, and which were included in-01 Report 84-006, dated March 7, 1984. However, the alleger was not able to identify specific plant locations involved or specific IR numbers or dates. The TRT then reviewed approxiniately two dozen irs prepared by one of the inspectors named, and approximately a dozen irs prepared by the other inspector named. The TRT could not determine from these records that the inspections were not actually performed. The TRT did not attempt further investigation of this allegation and considered it to be indeterminate with respect to technical significance. In addition, the TRT did not investigate further with respect to possible wrongdoing, which is beyond the scope of the TRT's responsibilities. (6) Rust Seen Through Tooke Gauge Tests (AQO-46). (a) It is alleged that, during Tooke gauge tests, rust was seen on steel substrate and grease, grime, filth, and other contaminants were observed on concrete substrate. (b) It.is alleged that Elcometer adhesion dollies, after being pulled off a coated surface, had rust adhering to the underside. It is also alleged that the QC Lead Inspector was aware of this condition and failed to take any corrective action. The TRT review of part (a) of this allegation indicated that it was not realistic for anyone to observe rust, grease, grime, filth, or other contaminants under 5 to 6 mils of primer and finish coat through a Tooke DFT gauge. An experiment conducted by.the TRT and a TUEC representative with a primer coat'of Dimecote 6 on-steel substrate. illustrated that Dimecote 6 could have possibly been mistaken as a light shade of rust, as it has a light, reddish gray color. A QC inspector without a great deal of experience with Dimecote 6 could mistake the reddish pigmenta-tion for rust. The TRT found no irs or NCRs that document this allegation. Coatings Category 7, " Training," assesses allegation AQ0-33; which was raised by another alleger and which relates to coating failure due to rust on the seal table A-frame steel. However, the A-frame rust was not established by a Tooke DFT gauge. The TRT review of part (b) of this allegation consisted of interviewing six QC inspectors certified for backfit test program inspections. Each of the inspectors stated that he had seen or heard of reports of adhesion tests which revealed surface rusting. However, no specific-locations were, indicated. Comanche Peak SSER 9 M-39 w__ _ _ ___
The inspectors stated their belief that, in each instance, the affected area was identified and repaired in accordance with procedural requirements. The TRT made no attempt to verify such repairs, since specific locations were not known. The signifi-cance of this allegation, to the extent that it is substantiated, is best judged by the backfit adhesion test results presented in Section 4.b, above. (7) Adhesion Tests Performed in Violation of Written Instructions (AQ0-47). It is alleged that, in violation of a written instructions, QC inspec-tors were instructed to perform approximately 25 Elcometer adhesion tests for an installation hanger for the steam generators. When a protective coating on steel fails an adhesion test, TUEC procedure QI-QP-ll.4-5 requires additional adhesion tests in the vicinity of the failed test to delineate the extent of unsatisfactory coating which must be repaired. Contrary to the allegation, the written instruction does not place an upper limit on the number of adhesion tests required to determine the extent of unsatisfactory coating. In response to this allegation, TUEC by letter from L. F. Fikar to NRC RIV, dated June 22, 1984, described a case where construction personnel requested QC inspectors to perform a total of 32 adhesion tests on the insulation support ring of steam generator No. 4 after failures were observed in the first two of three sets of adhesion tests. In this case, there were two failures in the last 26 adhe-sion tests, and the entire ring was stripped and recoated. It would have been a violation of Instruction QI-QP-ll.4-5 if areas which failed the adhesion test had not been reworked. In the absence of evidence that requests for additional pull tests were used to avoid reworking areas of unsatisfactory coatings, the TRT found this allegation to be unsubstantiated. (8) During the Backfit Test Program, Only the First Unsatisfactory DFT Reading was Recorded (AQO-54). It is alleged that during the backfit test program, only the first unsatisfactory reading was recorded, even if the following readings were either higher or lower, i.e. , further out of the acceptable range. It is also alleged that the trend analysis was adversely affected because the actual readings. were not included. The alleger's main concern appeared to be that if the first unsatis-factory reading of coating thickness was only slightly out of speci-fication, and this was the only unsatisfactory reading recorded for the sampled area, then the true thickness range for that area would not be known. If the range of acceptable thicknesses was later widened by a OCA to include the recorded reading, then the sampled area would be dispositioned as satisfactory even though other unre-corded thickness readings might have been outside the widened thick-ness specifications. Consequently, such an area would neither be repaired nor entered into the CEL. The specification changes widening the acceptable range of coating Comanche Peak SSER 9 M-40
thicknesses on steel and concrete are discussed in Section 4.b above, in the TRT's assessment of allegation AQ0-55 below, and in Coatings Category 2, "0BA Qualification Testing." The TRT was not able to ascertain independently the extent to which inspectors recorded only the first unsatisfactory thickness reading of coatings. However, the alleger stated that the problem was not widespread and that inspectors generally did not follow the verbal instruction of a particular QC supervisor to record only the first unsatisfactory reading unless the supervisor was watching. Other factors tending to mitigate the practical effect of the first-reading-only practice are the following. (1) Most of the out-of-specification thickness readings were too low. For the coatings used on steel at CPSES, adhesion is more strongly affected by primer or topcoat which is too thick. For the coatings which were used on concrete, the specified thickness range was much greater so that slight variations in the extent to which thickness was out-of-specification have a lesser effect. (2) Approximately 96 percent of the Containment Building liner and 50 percent of the concrete surfaces in Unit 1 were backfit-inspected for coating thickness. Except for a few areas of limited size, which were included in the CEL, all nonconforming areas of the liner and concrete were reworked until the coating thicknesses were in the allowable range. Only 22 percent of the miscellaneous steel coatings were backfit-inspected; the miscel-laneous steel items with nonconformir.g primer or topcoat thick-nesses were entered into the CEL by NCR-C-83-01305 instead of being repaired. i ( (3) According to TUEC QA personnel interviewed by the TRT, a large majority of the 22 percent of the backfit-tested miscellaneous steel items with low topcoat thicknesses were in fact reworked with more topcoat, partly for cosmetic reasons. A random sampling of irs by the TRT confirmed this. For similar cos-metic reasons, additional topcoat was applied to most of the coatings which had not been backfit-inspected (4 percent of the Containment Building liner, 50 percent of the concrete surfaces, and 78 percent of the miscellaneous steel surfaces). Because of the three mitigating factors and the alleger's statement that inspectors followed the first-reading-only practice only when under the supervisor's scrutiny, the TRT concludes that the practice would adversely affect the quality of only a small fraction of the protective coatings. l To look into possible wrongdoing relative to the verbal instructions j given by the QC supervisor as indicated above, the TRT Project Director forwarded the transcribed NRC interview with the alleger to NRC Region IV OI for its review in a letter dated August 24, 1984. (9) Areas Identified During the Backfit Test Program as Requiring Coatings Removal did not Have Coatings Removed (AQO-55). It is alleged that Comanche Peak SSER 9 M-41
areas identified during the backfit test program as being outside of
. the acceptable thickness range for applied coatings were not reworked
) as required. As discussed above in 4.b, the ranges of acceptable coating thick-nesses were widened by DCAs (e.g. OCA 12,145). The widened ranges were incorporated into the later revisions of procedures QI-QP-11.4-23 and QI-WP-11.4-24. After these revisions, some coatings which met the expanded specifications were dispositioned "use as is," whereas i repair would have been required by the earlier specifications. The TRT evaluation of allegation AQ0-54 stated that NCR-C-83-01305 i permitted miscellaneous steel items with nonconforming primer or 1 total coating thickness to be placed in the CEL rather than being repaired. TUEC procedures and NCRs required the repair of noncon-forming coating thicknesses on Containment Building liner and con-1 crete surfaces. In its review of inspection reports and NCRs related to protective coatings, the TRT found no instances where unsatisfac-3 tory coatings were dispositioned "use as is" without being entered into the CEL. However, according to interviews with TUEC QA personnel, the entry of miscellaneous steel items into the CEL (the first four l entries in the CEL) was based on a TUEC assumption that 5 percent of each category of miscellaneous steel was of indeterminate quality. As discussed in Section 4.b and in Coatings Category 6, " Coating Ex-empt Log," this assumption may be low.
, The uncertainty regarding the miscellaneous steel CEL entries may be ! considered to be a partial substantiation of the allegation. The TRT t found no other evidence confirming this allegation.
4 (10) BTP Occumentation Destroyed (AQ0-56). It is alleged that original documentation related to the backfit test program was destroyed by
; QA management.
This allegation involves possible wrongdoing and is outside the TRT's assigned scope; therefore the TRT did not directly investigate this allegation. In addition, since the alleger could not identify which i documents were destroyed, the TRT was unable to make any cross-check from related documents to substantiate the allegation, nor could the TRT make any technical assessment of the impact of the alleged destroyed documents on the quality of coatings. The alleger in-directly relates this allegation to forging and falsification of BTP documents (alleged by another person) which was assessed in allega- , tion AQ0-37(c). In a letter dated August 24, 1984, the TRT director forwarded the i transcribed NRC interview with this alleger to NRC Region IV OI for its review.
- 5. Conclusions and Staff Position:
- a. Specific Allegations Concerning BTP. The TRT's conclusions concerning specific allegations related directly or indirectly to the coatings backfit test program (BTP) are as follows:
! Comanche. Peak SSER 9 M-42 i 4 l
(1) Allegations AQ0-18, -20, and -37(a) were substantiated; however, for I the reasons stated in the corresponding assessments, the TRT does not consider them to be of significant technical concern in regard to the BTP. Allegation AQ0-21 was also substantiated. As discussed in detail in 4(b), the TRT concludes that correcting adhesion test results for Elcometer calibration error has only a small effect on liner and concrete test failure rates, but could significantly change the test failure rate for miscellaneous steel. For the liner plate, the fail-ure rate changed only from TUEC's uncorrected 2.3 percent (20/869) to the TRT's corrected rate of 4.3 percent (36/834) or to TUEC's corrected rate of 5.9 percent (51/869). For the concrete, the change was from a 0 percent uncorrected failure rate to the TRT's corrected 6.8 percent (8/116) rate or to TUEC's corrected rate of 3.1 percent (65/2128). For miscellaneous steel, however, the test failure rate could change from less than 1 percent (26/4714), uncorrected, to as much as 19 percent corrected, based on a limited sample analysis by the TRT. (2) Allegations AQ0-37(c), -54, and -55 were partially substantiated. The TRT could not draw a conclusion as to forging or falsification of documents for allegation AQ0-37(c), but did find two instances of tech-nically inconsistent inspection documents which may involve only a small area of liner coating. The TRT does not consider allegations AQ0-54 and -55 to have any significant effect since the TRT believes that the area which was not reworked due to coating thickness out of tolerance is small. However, TUEC's estimated 5 percent miscellaneous steel coating area allowance due to indeterminate quality may be low. (3) The TRT concludes that allegations AQ0-19, -37(b), -46(a), -46(b), and 47 were not substantiated, and therefore are of no concern. (4) The TRT concludes that allegation AQO-37(d) was indeterminate with respect to technical significance. The TRT did not investigate possible wrongdoing in regard to this allegation. (5) The TRT did not investigate allegation AQ0-56 since it involves wrong-doing and there was insufficient information available to assess its potential extent and effect on coating quality. This allegation has been referred to NRC Region IV OI for review. Even though the TRT found that most of the allegations were of small technical significance with respect to the BTP overall pass / fail rate, many of them, as indicated in the individual assessments, relate directly to TRT generic concerns about other aspects of the TUEC coating QA program and are addressed in other Coatings categories,
- b. Evaluation of Backfit Program Test Results. The TRT evaluated the BTP inspection and test field data to determine independently, within practical limits, the extent of the coated areas tested by TUEC and the extent to which tested coatings passed their DFT and adhesion test acceptance criteria.
Comanche Peak SSER 9 M-43
I I It is the TRT's opinion that if DBA qualified coating was applied l and its traceability was maintained, then the BTP, properly admin- ! istered, could provide useful, but indirect information on the ' quality and DBA survivability of the coating work. However, the TRT found serious shortcomings in DBA qualification and traceability of l' coatings as discussed in Coatings Categories 2 and 3, respectively. Further, the TRT does not consider that the BTP scope or results provide information that resolves these shortcomings. Nevertheless, the TRT concludes that the thousands of tests conducted under the BTP provide a useful overall measure of .two important coating quality parameters: adhesion strength and coating thickness. The results of the BTP adhesion tests, corrected for Elcometer calibration error, are summarized in the following table. On the ! basis of these results, reasonably accurate estimates of the adhesion ! test failure rate and the corresponding failed areas can be made for the coatings on the Containment Building liner and concrete surfaces. For the miscellaneous steel surfaces, the adhesion test failure rate could be only crudely approximated, because corrected adhesion test ' data are lacking except for the small sample audited by the TRT. The TRT's corrected data failure rate is approximately.19 percent, in contrast to TUEC's uncorrected data failure rate of less than 1 per-cent. A 19 percent miscellaneous steel failure rate corresponds to a failed coating area of 34,200 square feet. As shown in the table, this figure, in addition to the failed areas of the Containment Build-ing liner and the concrete, gives a total of approximately 57,500 square feet of coatings which failed adhesion tests. Currently, less than 3000 square feet are entered into the CEL for this reason. (See Coatings Category 6.) In addition to the adhesion tests, the other BTP test performed was for coating thickness measurement. For the liner and concrete, coatings with thicknesses outside of the allowable range were re-worked until the thicknesses were acceptable (although the allowable range was expanded several times) in nearly all cases. In the low percentage of cases, where "epair was not feasible, the discrepant areas were placed in the CEL. - For miscellaneous steel with unsatisfactory coating thicknesses, the item could be dispusitioned "use-as-is" and placed in the CEL by NCRs, e.g., NCR C-F,3-03103. Rather than attempt to estimate the area of each discrepant item, TUEC conservatively entered approximately 5 percent of each type of miscellaneous steel area totalling 8150 square feet in the CEL. The 5 percent entry may be low because the DFT test failure rate was 8.5 percent cosmetic rework. In summary, based on its review of BTP data, the TRT estimates that 90 percent of all coated surfaces meet adhesion test requirements (assuming a 19 percent failure rate for miscellaneous steel). The TRT estimates that more than 90 percent of the coated surface thick-ness was acceptable (allowing for repeated relaxation in thickness tolerances). However, the extent to which the BTP demonstrated Comanche Peak SSER 9 M-44
..q n ADHESION TEST DATA
. E l I h TRT ESTIMATE TUEC ESTIMATE y Total Area % Area Failure Failed % Area failure Est. Failed
- g. Item ft2 Audited Rate, % Area, ft2 Tested Rate, % Area, ft2 M
l g Containment 110,000 79% 3.6% 3960 96% 5.9% 6490 l . Liner l Concrete 285,000 3.9% 6.8% 19400 50% 3.1% 8835 Miscellaneous 180,000 0.36% 19% 34200 22% Not Not Steel Available Available x
$ Total 575,000 57560
i I . that FSAR coating quality requirements were met is dependent on resolution of other TRT generic concerns related to DBA qualification and traceability of coatings, as discussed in Coating Categories 2 and
- 3. These deficiencies, although not of safety significance in the 7
- coatings area, must be considered in evaluating the effectiveness of 1 TUEC's overall QA/QC program.
For an integrated assessment of the estimated total coating area inside the Containment Building which failed the BTP and/or had indeterminate quality for other reasons, including vendor equipment with unqualified coatings, refer to Coatings Category 6, " Coating Exempt Log." j Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original concerns and to obtain any additional comments from them. A summary.of the followup interviews is included in Section 2.2.3 of this Appendix. j 6. Action Required: The TRT reviewed the procedures used and the inspection j reports (PCRs) and statistical data which resulted from the Coatings i Backfit Test Program conducted for coatings applied to the Containment Building liner, concrete structures, and miscellaneous steel. i For the 2189 miscellaneous steel items (such as pipe hangers, cable trays, j equipment, and conduit supports) that were tested under the backfit test
- program from about December 1981 to February 1984, the TRT found that TUEC
] did not correct any of the coating adhesion test field data to account for 4 the calibration error of the Elcometers used for the tests. Consequently, the field test data could be in error by as much as 200 psi in a noncon-servative direction. Thus, any Elcometer reading less than 400 psi potentially represents a test that failed to meet the test acceptance criterion of 200 psi. I Based on TRT analysis of sampled data covering 78 test results for 42
- miscellaneous steel items, the appropriate calibration correction could
! increase the adhesion test failure rate from about 1 percent to 19 percent. ! The following inspection reports (PCRs) indicate acceptable test results ! (>200 psi) before correction, but unacceptable test results (<200 psi) after correction: PCR-02103 PCR-02164 PCR-02105 PCR-02166 PCR-02515 PCR-02171 ! Accordingly, TUEC shall: 1 (1) Apply the Elcometer calibration correction to the data for the 4714 adhesion tests covering 2189 miscellaneous steel items tested to establish a more reliable estimate of the adhesion test failure rate. q This revised analysis should include a statistical analysis showing i the 95 percent confidence upper limit of the failure rate for all the j miscellaneous steel items inside the Containment Building. I Comanche Peak SSER 9 M-46
(2) Analyze the corrected data to establish a more reliable estimate of the fraction of tested miscellaneous steel coated surface that failed the adhesion test acceptance criterion. The resulting failed areas shall be entered in the Protective Coating Exempt Log. (See Coatings Category 6.) The coatings exempt log will be used in planning future inspections of coatings consistent with the guidelines of Appendix L. l Comanche Peak SSER 9 M-47
i
- 1. Allegation Category: Coatings 2, Design Basis Accident (DBA) Qualifica-tion Testing
- 2. Allegation Number: Parts of AQ0-01, AQO-02, AQ0-03, AQ0-04, AQO-05, AQ0-06, A00-09, AQ0-10, A00-11, AQ0-12, AQ0-13, and AQ0-15
- 3. Characterization: It is alleged that some protective coating systems applied at the Comanche Peak Steam Electric Station (CPSES) are not DBA qualified. Examples are:
B&R Procedure CCP-40, Paragraph 4.3.1.2, allows the application of i Imperial coatings (Southern Imperial Coatings of New Orleans) in the sequential order 115/1201/115/1201 or 115/1201/11/1201. The alleger questions whether adequate testing has been performed to demonstrate the 08A qualifications of these materials applied in these sequences (AQO-01). Nonconformance report (NCR) C83-01752, June 23, 1983, indicates that repair _ coating systems can be applied in sequences different from the original application sequences. The alleger questions whether adequate testing has been performed to demonstrate the DBA qualifica-g tion of materials applied in these different sequences (AQ0-02). Design change authorization (DCA).17,142, Revision 2, permits 1 Carboline Phenoline 305 (P-305) to be applied over another manufac-turer's epoxy coating. The alleger questions whether this combina-tion of different manufacturers' coatings has been DBA tested (AQO-03). l DCA 12,374, Revision 1, permits Carboline Carbo Zinc 11 (CZ-11) to t be topcoated with Imperial 1201. The alleger questions whether this I combination of different manufacturers' coatings has been DBA tested (AQ0-04). CPSES Procedure CCP-30A, Revision 2, Paragraph 1.3.1, allows P-305 to be applied over Ameron Dimecote 6 (D-6). The alleger questions whether this combination of different manufacturers' coating mate-erials has been DBA qualification tested (AQO-05). l CPSES Procedure CCP-40, Revision 5, Paragraph 4.1.1.3, permits Imperial Nutec 115 surfacer to be applied over foreign objects embedded in concrete. The alleger questions whether adequate i testing has been performed to demonstrate the DBA qualifications of this material applied over foreign objects (AQ0-06). I - It is alleged that inorganic zinc primer has been applied in three coats and that this three-coat application method has not been DBA qualification tested (AQ0-09). It is a'lleged that coatings applied to surfaces which were prepared by power tool cleaning were smooth'ed or polished, and thus do not have adequate surface profile. The alleger questions whether adequate testing has been performed to demonstrate the DBA qualifications of materials applied to such smoothed or polisned surfaces (AQ0-10). Comanche Peak SSER 9 M-49 i l
It is alleged that DCA 18,489, allows the application of primer at a thickness of 0.5 mils. The alleger questions whether DBA testing has been performed for primer this thin (AQ0-11). CPSES Procedure CCP-40, Revision 5, Paragraph 4.3.1.2, allows applica-tion of the Imperial coating system at a thickness of 102 mils. The alleger questions whether this system has been DBA qualification tested for coatings this thick (AQ0-12). It is alleged that adequate testing has not been performed to demon-strate the DBA qualifications of coatings applied in the reactor core cavity where they will be subjec?.ed to higher levels of neutron and gamma exposure than coatings in other areas (AQ0-13). CPSES Procedure CCP-30, Revision ll, Paragraph 4.4.3.0, allows the application of CZ-11 or Carboline 191 primer (191P) over P-305, and the application of P-305 over Imperial 1201. The alleger questions whetbar adequate testing has been performed to demonstrate the DBA qualifications of these materials applied in these non-standard combinations and sequences (AQ0-15).
- 4. Assessment of Safety Significance: The majority of allegations in this category address botn OBA qualifications and procedures which may have permitted the application of non-DBA qualified coatings. The procedural aspects of these allegations are addressed in Category 4.
Review of DBA Qualification Test Data In assessing these allegations, the NRC Technical Review Team (TRT) reviewed the requirements pertaining to DBA testing of protective coating materials used inside primary containment structures at CPSES. The CPSES FSAR, Section 6.1.2, and Technical Specification AS-31 commit Texas Utilities Electric Company (TUEC) to compliance with American National Standards Institute (ANSI) standards N101.2 and N512, which provide the methods and criteria for DBA testing. This testing must be performed under temperature, pressure, and irradiation conditions which simulate both normal plant operating conditions and loss of coolant accident (LOCA), i.e., DBA, conditions. These standards specify that the coating systems tested must be the same as those actually used in the plant in terms of surface preparation, coating materials, application methods and conditions, and coating thickness. CPSES Specification 2323 AS-31, which governs safety-related coating work, requires that the D3A testing be performed by both the coating manufacturers and Oak Ridge National Laboratories (ORNL). The CPSES FSAR, Section 6.1.2, also commits TUEC to compliance with ANSI N45.2.11, which provides require-ments for design activities, includina evaluation of design input informa-tion (such as DBA tests), both for the initial design and all changes thereto, and for records which must be maintained to document design activities. The TRT investigated compliance with these requirements at CPSES, con-centrating on the coating systems which were used for the major surface Comanche Peak SSER 9 M-50
areas inside the Containment Building. Secondary consideration was given by the TRT to the many different " coating systems" that were used in cer-tain small areas, such as overlaps between different systems, repairs, and other special cases, including those described in the allegations. The TRT investigation of the DBA concerns included interviews with approxi-mately 12 individuals involved in design and engineering relative to coatings at CPSES. Those interviewed were from Gibbs & Hill, ORNL, Texas Utilities Service, Inc. (TUSI), TUGCO, EBASCO, and Brown & Root and were involved in design, chemical engineering, DBA testing, field engineering, design changes, design and design change control, records management, and The TRT conducted an in-depth review of numerous DBA quality assurance.and other test reports, of related documentsThe provided TRT also by TUEC, of pertinen CPSES site files of OCAs, and of vendor-approval records. reviewed relevant American National Standards Institute (ANSI) industry standards, NRC information reports, TUEC licensing documents, TUGC0 procedures, and Regulatory Guides. The TR1 requested from TUEC copies of DBA test data and On evaluations appli-August 10, 1984, cable to CPSES containment structure coatings systems. TUEC provided a package labeled "DBA Reports - Site Civil Engr File," and on August 29, 1984, TUEC provided five packages of DBA data from the CPSES vault. TUEC explained that the five " vault packages" were the official DBA Test Reports which demonstrate the acceptability of the major coating systems at CPSES, and this was confirmed in a September 3, 1984, TUGC0 memorandum from the former CPSES QA Manager to the TUGC0 Vice President and CPSES Project General Manager, a copy of which TUEC presented to the TRT on September 10, 1984. The TRT reviewed all of the above test reports and found them to be incomplete, inconsistent, and inadequate in demon-strating that the major coating systems usedTheat CPSES TRT verified had been this qualified finding by in accordance with ANSI N512 and N101.2. performing a second review of only the five official vault packages that Although the packages contained some pertinent yielded the same results. testing data, they also contained a large quantity of in was not applicable or not valid, and the pertinent data was not adequate to demonstrate that the coatings were qualified. The TRT found no records to demonstrate that the DBA testing data had been reviewed by CPSES personnel and no records of engineering evaluation of the data by TUEC, either in the test reportInformation packages,provided the vendor file, to the TRTor elsewhere, as required by ANSI.N45.2.11. by TUEC in the September 3, 1984, memorandum indicated that Gibbs & Hill (the CPSES Architect / Engineer) was involved in the design process for coat-ings, but did not demonstrate that an engineering evaluation of the DBA test data had been performed. The TRT concluded that such a recorded engineering evaluation was necessary, not only because it is required by ANSI N45.2.11, but also because the files of testing data by themselves do not demonstrate the qualifications of the coatings without interpreta-tion and additional data. The TRT observed that the DBA data base files were deficient in that: M-51 Comanche Peak SSER 9
(a) There were three different test report data bases, each containing different data: those used by Gibbs & Hill; those used by CPSES field engineering; and those kept in the CPSES vault. (b) Many of the reported DBA tests did not include all important criteria (temperature, pressure, and irradiation). (c) Many of the reported DBA tests were not performed by Oak Ridge National Laboratories (ORNL), as required by CPSES specification 2323-AS-31. (d) Many of the reported tests included coating materials and coating systems not used at CPSES in combination with materials and systems which were used at CPSES. The inclusion of these " foreign" materials in the systems tested invalidates the applicability of this test data to CPSES. (e) In some of the reported tests, the test samples failed, indicating that these materials and systems were not qualified to the required criteria, yet these reports were included in the qualification test data base without comment. (f) Many of the test reports did not include sufficient data to identify either the materials or system tested, the test criteria, or the results. (g) The data bases included instances of two different, incomplete. versions of the same report. (h) The data bases included an instance of two different reports of two different tests of the same system showing different results, i.e., one report documents that the test was passed and a second report documents that the same system failed the test. (i) The data bases did not include any one complete, acceptable, valid test that corresponded to each major CPSES coating system. The following two examples illustrate the deficiencies observed in the test data files: (1) The site field engineering files included Carboline test report SR-149 that recorded an ORNL test report of August 16, 1978, which i l included temperature and pressure only, for the following system: CZ-11, at 1.7 to 2.9 mils, topcoated with P-305 at 1.5 to 3.8 mils. In the official vault file there was a different version of this same 1 ORNL report of August 16,'1978, which included temperature and pres-sure only for the following system: CZ-11, at 3 to 4 mils, with a topcoat different from that used at CPSES. (2) The official vault package included a Carboline Laboratory test report The report gave for Testing Project Number 01931, February 10, 1981. the results of a test performed by Carboline for temperature and pressure criteria only on the following repair system: Carboline 191 Comanche Peak SSER 9 M-52
I primer (191P) at 4.0 to 4.5 mils, topcoated with P-305 at 3.0 to i 3.5 mils, over a power-tool-cleaned surf ace (that had previcusly been abrasive blasted and coated with the CZ-11/P-305 system). The report indicates that the samples tested passed. The site field engineering package included a report of Carboline Testing Project 01978 which reported a DBA test by ORNL of April 7, 1982, which included tempera-ture, pressure, and irradiation for the same system: power tool cleaning, 191P at 4.5 to 5.25 mils, and P-305 at 3.25 to 4.0 mils. In this report, two of the six samples tested at ORNL failed the test. As a final check on the validity of the DBA test data, the TRT analyzed the data in the five official vault packages for the major steel coating system at CPSES: abrasive blasting to Steel Structures Painting Council (SSPC)
! standard SSPC-SP-10 (Near White Metal), CZ-11 primer, and P-305 topcoat.
The dry film thickness acceptance criteria for this system at CPSES (as provided by CPSES procedures for coating work) is 1.5 to 7.0 mils for CZ-11 and P-305 thickness adequate to provide " full hiding," but with a total system thickness less than 15.0 mils. The purpose of the TRT analy-sis was to determine whether these packages included enough valid informa-tion to enable an engineering evaluation to be performed to demonstrate that this system, as applied at CPSES, meets the criteria of ANSI N101.2 and N512. For the analysis, the TRT assumed the accuracy of the data .in the packages and the technical applicability of the testing performed by the manufacturer, although such testing did not include irradiation. Test data were not considered in this analysis if the test did not include either CZ-11 or P-305, if the sample (s) tested did not meet the ANSI N101.2 and N512 acceptance criteria, or if the test report did not include suf-ficient detail to identify the particulars (e.g. , dry film thickness) of the material (s) tested. The TRT found that: (a) the only item for which j there was full ORNL testing data was CZ-11, by itself, at 2.0 to 3.5 mils
; 0FT; (b) there wa's no irradiation data for P-305 by itself or for the i CZ-11/P-305 system; (c) there was no independent (ORNL) testing data for the CZ-11/P-305 system, as a system; and (d) there was no testing data for the CZ-11/P-305 system that included or closely approached the thickness ranges used at CPSES. The TRT concluded that this testing data did not l provide adequate information to enable an engineering analysis to be 1
performed which would demonstrate the qualification of this, coating system
- as applied at CPSES.
In light of-the deficiencies in the test data files, and particularly because there were instances of different incomplete versions of the same ORNL report and different test results for the same system, the TRT consi-dered whether TUEC had taken any actions in response to NRC IE Informa-tion Notice No. 83-60, (This notice reports an incident of falsi.fied DBA test reports by a coatings vendor and recommends that NRC applicants and licensees can avoid such incidents by obtaining coatings test reports directly from independent testing laboratories.) The TRT found that TUEC had received and analyzed this notice and determined Oat no action by them was appropriate because they did not require their coatings to be tested by an independent laboratory. The TRT found that'all testing reports were, in fact, provided to TUEC by the coating manufacturers. Although it is i true that the NRC did not require any' action by TUEC based on the IE Notice, the TRT disagrees with TUEC's analysis and conclusion that no action was Comanche Peak SSER 9 M-53 i
appropriate, because that conclusion was based on the erroneous statement that TUEC's procedures do not require independent laboratory testing. CPSES Specification AS-31 does, in fact, require DBA testing by an independent laboratory, in this case ORNL, as well as by the coating manufacturer. The TRT also examined vendor files to determine if they contained evidence of engineering evaluation of DBA test data as part of the vendor approval process. This examination included a review of vendor files for coating manufacturers Ameron, Carboline, and Imperial. These files did contain evidence of a functioning vendor-approvai system which included evaluation of vendor QA program manuals, surveillance and audits at the vendors' facilities by TUEC representatives, and inclusion of pertinent quality requirements in purchase orders to the vendors. However, these files did not include evidence that TUEC engineering had evaluated DBA test data. Review of Design Changes The TRT next investigated TUEC's design change process to learn whether engineering evaluation of DBA test data was properly used asinvestigation This justification for design changes that affected CPSES coating systems. included interviews with personnel involved in the design change process, a review of TUSI procedures CP-EP-4.0 and CP-EP-4.6, and an examination of coatings design change records. (The subject of design changes is addressed in further detail in Coatings Category Sc.) The TRT reviewed all revisions of the major instructions which governed coatings work at CPSES, including Procedures CCP-30, CCP-30A, and CCP-40; selected inspection procedures, including those for backfit inspections; specification 2323-AS-31; and design changes which affected the above. (Further details on procedures and on the backf.it program are provided in Coatings Categories 1 and 4.) The TRT found that there were many substan-tive changes in the coating systems as applied, including changes in allow-able thickness criteria, coating materials, sequencing and combinations of materials, surface preparation and coating application methods, and accept-ance criteria. All such changes affect the coating system design in that they affect the relationship between the coatings as applied at CPSES and the coatings as DBA tested. The TRT evaluated design changes affecting coating work which were selected from those identified in allegations and those found by the TRT during its review of TUEC's procedures and instructions for coating work. Eight design changes were chosen because they involved changes for which engineering evaluation and justifi. cation would be necessary to demonstrate the DBA qualifications of the resulting coating systems. Design Change Authoriza-tions (DCAs), Change Verification Checklists, and supporting documentation were reviewed for each. In some cases the engineering justification for approval of the change was evident in the documents reviewed, e.g., cases. where specification requirements were not met and the coated item in ques-tion was placed in the coatings exempt log. However, the majority of the documentation reviewed did not provide any engineering justification or other reason for approval of the change, such as evaluation of new DBA test data. The TRT concluded that records were not adequate to demonstrate Comanche Peak SSER 9 M-54
the engineering basis for design changes for coatings at the time those changes were approved. In response to the TRT's further inquiries, TUEC developed written explana-tions of the engineering justification for certain design changes and pro-vided them to the TRT in the form of an attachment to the September 3, 1984, memorandum (from the former QA Manager to the TUGC0 Vice President and Project General Manager). These explanations reference manufacturer's test reports, ORNL test reports, and accepted industry standards as the basis for the changes. The TRT does not fully agree with the technical validity of these explanations in all cases. For example, in the explan-ation, they apply a standard which gives a range of allowable deviations from a specified single point coating thickness, to a pre-specified thick-ness range. The TRT does not consider this an appropriate application of this standard. Their explanation also cites a report of testing performed by Carboline, which does not include an irradiation test, for various thicknesses of CZ-11 topcoated with 1 mil of P-305, as justification for increasing the thickness of CZ-11 with a much higher thickness of P-305 topcoat. The TRT considers that this test report is not applicable without additional supporting information and evaluation, neither of which were provided. Review of I_ndivi_ dual A_ll_egations The TRT evaluated the allegations that certain coating systems were not DBA qualified by investigating the DBA qualifications of the various coating systems, which were used for touchup work, overlapping, and other special cases described in the allegations. Imperial Concrete Coating Sequences (AQ0-01). The following sequences of Imperial concrete coatings are permitted by CPSES Procedure CCP-40, Paragraph 4.3.1.2: 115/1201/115/1201 and 11S/1201/11/1201. The TRT reviewed CCP-40, Imperial Letter VBR-7697, May 8, 1978, Imperial Technical Report 759, April 19, 1984, Imperial Technical Report 495-81, June 10, 1981, and related correspondence from TUEC to the TRT. CCP-40 permits application in these sequences for repair and touchup work. Imperial Letter VBR-7697 states, "Although the resultant systems 115/1201/11S/1201 or 115/1201/11/1201 have not been qualification tested, there is no reason to believe that they are not viable systems." It is clear that these systems were used and that they were not qualified, a condition which TUEC has acknowledged in correspondence to the TRT. 4 Imperial Technical Report 759 provides some information to support a find-ing that these systems are technically viable. This report documents in-house adhesion tests of these systems performed by Imperial, with accept-able results. Imperial Technical Report 495-81 provides OD.NL test data that indicates that the individual coating materials in these systems will perform satisfactorily under normal service and DBA conditions. TUEC stated in correspondence to the TRT that these systems were used only in limited areas as needec for overlapping to achieve a smooth application adjacent to previously coated surfaces and repairs. Because of.the areas where the overlap was used, TUEC stated that information on the total sur-face area is not available, and that these areas have not been entered in the coatings exempt log. During examinations of coated concrete surfaces Comanche Peak SSER 9 M-55
at CPSES, the TRT did not observe any noticeably thicker areas or ridges in the concrete coatings or any other evidence to refute TUEC's statement that N1y limited areas of smooth transitions were involved. The TRT concluded that although these systems are not DBA qualified, their use at CPSES was acceptable. Repair System Sequences Which Differ _From Original _ Sequences (AQ0-02). The TRT reviewed CPSES NCR C83-01752, June 23, 1983, Table A2 in Appendix A of Specification 2323-AS-31, CPSES Procedures CCP-30, Rev. 10, January 26, 1982, CCP-30A, Rev. 2, September 20, 1982, and CCP-40, Rev. 5, August 18, 1982, and related correspondence from TUEC to the TRT. In NCR C83-01752, the disposition states, " Table A2 in Appendix A of AS31 specifies acceptable coating systems, i.e., primer and final coat, product identification, and vendors . . . . This table does not identify full system sequencing or application parameters." The TRT's review of Table A2 confirmed the accuracy of these statements; however, a note to this table found in Rev. 2, March 15, 1984, of specification AS-31 states, "It is essential that coating systems be used only as specified above, unless an alternate system is proposed by a coating manufacturer and subsequently approved by the engineer." The concern expressed in this allegation is that repair coatings are applied in different sequences than the original coating system and that these repair sequences are not DBA qualified. The TRT's review of CPSES application procedures CCP-30, CCP-30A, and CCP-40, generally supports TUEC's statement in correspondence to the TRT that repair sequences are not different from original application sequences, except in areas where they overlap. (Coatings Category 4 provides additional information about these procedures and an assessment of their adequacy. The adequacy of DBA test data for the original application sequences, and engineering evalua-tions of changes to them, are discussed under Review of DBA Qualification Test Data in this assessment.) TUEC stated in correspondence to the TRT that nonstandard sequences occurred only in areas of overlap, that the total area involved is indeterminate but minor, that these areas are not significant enough to warrant special consideration, such as DBA qualifi~ca-tion, and that they are not included in the exempt log. Except as specifically noted elsewhere in this assessment, the TRT con-cluded that applying coatings in nonstandard sequences that were not spe-cifically DBA qualified was acceptable when used in small areas of repair and overlap. Applying P-305 Over Another Manufacturer's Epoxy Coating GQC-03). In assessing this allegation, the TRT reviewed correspondence from TUEC on this subject, TUSI Procedure CP-EP-16.4, Rev. O, October 31, 1983,
" Protective Coatings Exemption Log," selected entries from the Coatings Exempt Log (CEL), and DCA 17,142, Revisions 2 and 3. (Further discussion of the CEL is provided in Coatings Category 6.) The TRT found that although Revision 2 of DCA 17,142 permitted the use of this nonqualified system, Revision 3 provided justification by reference to entry No. 22 in the CEL. The TRT's rev.iew of the log found that entry No. 22 recorded Comanche Peak SSER 9 M-56
l l 2300 square feet of P-305 applied to the Unit 1 manipulator crane. This area is considered small compared with the approximately 600,000 square feet of coated surface inside the Unit 1 Containment Building. Because entry of an item into the CEL effectively removes it from Service Level I and Specification 2323-AS-31 requirements, the TRT concluded that this item had been handled properly by TUEC and was without adverse consequences. CZ-11 Topcoated With 1201 (AQ0-04). The TRT reviewed DCA 12,374, Rev. 1, November 2,1982, CEL entry No. 30, Imperial Technical Report #553-81, and related correspondence from TUEC to the TRT. The TRT found that this version of Imperial Technical Report 553-81 (which consisted of extracts from a larger report) gave enough information to provide confidence that the CZ-11/1201 system met DBA qualification requirements. That report, however, does not provide sufficient detail to identify the particulars of the system tested; for example, application methods and conditions and dry film thickness (DFT). The TRT found that DCA 12,374 provides for the coating of Richmond inserts in concrete using this system in accordance with the requirements of CPSES Specification 2323-AS-30, which is the nonsafety-related coatings specification. The DCA does not mention or reference either test report 553-81 or the CEL as justification for this change. The adequacy of the DBA test report data base and of recorded engineering evaluations of design changes is discussed under Review of DBA Qual _ification Test _ Data elsewhere in this assessment. The TRT reviewed the CEL and found that entry No. 30 records 2258 square feet of this system applied to Richmond inserts in the Unit 1 Containment Building. The TRT found that the area estimated was acceptably conserva-tive. (Coatings Category 6 provides additional discussion on the CEL.) Because this item is included in the CEL and involves a relatively small area, the TRT concluded that it is without adverse consequences or safety significance. P-305 Applied Over D-6 (AQ0-05). The TRT reviewed CPSES Procedure CCP-30A, Rev. 2, September 10, 1982, " Coating Steel Substrates Inside Reactor Build-ing and Radiation Areas," Carboline Testing Project 01684, August 11, 1978, and related correspondence from TUEC to the TRT. The TRT found that Pro-cedure CCP-30A is the procedure for application of alternate coating system D-6/P-305 to safety-related steel surfaces. Revision 2 of this procedure provides for a DFT of 1.5 to 5.5 mils for D-6, with total system thickness of 7 to 11.5 mils. TUEC provided the Carboline Testing Project 01684 report to demonstrate the DBA qualification of this system. The TRT reviewed the report and found that it documented a test performed by Carboline for this system, which ir.cluded temperature and pressure criteria only, for thick-ness ranges of from 2.4 to 3.5 mils for D-6 and from 7.3 to 11.1 mils for the total system DFT. The TRT has little technical concern that this system will not perform adequately in service. This is because, in the TRT's opinion, the use of Ameron D-6 with Carboline P-305 is technically acceptable, in that both materials are commonly used nuclear safety-related materials. Ameron D-6 is an inorganic zinc, generically similar to CZ-11, and is commonly used at other nuclear facilities with various Ameron epoxy top coats which are generically similar to P-305. However, the TRT notes that the testing Comanche Peak SSER 9 M-57
data provided by TUEC do not demonstrate that this system is qualified, since no irradiation data are included for either material or for the system; other related test data for these materials as used in other systems have not been provided, referenced, or evaluated; and the Carboline test report is not independent and does not include testing over a broad enough range of thicknesses to provide comparability to the thickness range used at CPSES. 115 Applied Over Foreign Objects ( AQ0_-06 ) . This allegation concerns DBA qualification of Imperial Nutec 11S surfacer applied over various foreign objects, such as nails, rebar chairs, bolts, wood, or plastic, embedded in concrete. The TRT reviewed CPSES Procedure No. CCP-40, Revision 5, August 18, 1982, " Protective Coating of Concrete Surfaces," correspondence between Imperial and TUEC, correspondence from TUEC to the TRT, and Imperial Technical Report No. 462-81, January 22, 1982, which was provided with TUEC's correspondence to the TRT. The TRT found that TUEC had investigated this item to determine the tech-nical consequences of applying NUTEC 115 over embedded objects. The TRT also found that there was no DBA test data to demonstrate the qualifica-tion of such applications. The TRT reviewed Imperial Technical Report No. 462-1-81 and noted that it included ORNL testing of NUTEC 115 over steel in which the tested samples did not meet the ANSI N101.2 and N512 acceptance criteria. The fact that, in correspondence to the TRT, TUEC stated that "These tests demonstrate adequate coating performance under DBA conditions" is illustrative of inadequate engineering evaluation by , TUEC. Because of the varied locations of embedded objects, TUEC stated - that information concerning the total surface area involved was not available, and that these areas were not included in the CEL. t The TRT observed that, except as noted above, TUEC ma'de reasonable efforts ! to determine the proper treatment of embedded foreign objects and to perform affected coating work in accordance with the coating material manufacturer's recommendation. Because the areas involved were relatively small, the TRT found that although these systems are not DBA qualified, ; TUEC's treatment of this item was acceptable. I Application of Inorganic Zinc Primer in Three Coats (AQO-09). This allegation is concerned with the application er three coats of inorganic zinc primer at CPSES, which was alleged to be contrary to CPSES Procedure
- No. QI-QP-11.4.5, Revision 27. It-is also alleged that a three-coat system would lack adequate cohesion and.is not DBA qualified. ,
The TRT reviewed Procedure QI-QP-11.4.5, CPSES Procedures CCP-30 and CCP-30A, correspondence from TUEC to the TRT on this subject, and DBA
- and related test ~ data provided by TUEC, including Carboline Laboratory Test Report No. 01978.1, May 14, 1982, Carboline Testing Project 02182, ,
October 27, 1983, and Ameron Test Report No. TRC-089-03, October 10, 1979. Both the TRT and TUEC noted that the application of three coats of , inorganic zinc is not a violation of CPSES procedures. The test reports ' provided adequate data to demonstrate that application of inorganic zinc , in two coats rather than one would have no adverse effect upon material ; qualification. Furthermore, the TRT determined that excessive DFT due ; Comanche Peak SSER 9 M-58 -_____________r-________-____-____-____-_--__- _- - . _ - _ _ _ - _ . _ . - -
to multiple coat application was not a problem because overcoating was employed for the specific purpose of bringing thin areas up to the required minimum thickness. Accordingly, no cohesion or compatibility problem is caused by multiple coats of inorganic zinc, provided that proper thinning proce :ures are followed and that the total dry film thickness is within ace ptable limits. The TRT observed that proper thinning practices for overcoating inorganic zinc with inorganic zinc were in effect onsite. This allegation was not substantiated. Coatings Applied Over Surfaces Cleaned by Power Tools (AQO-10). This allegation concerns the UBA qualifications of repair or touch-up coating work in limited areas in which coatings are applied over surfaces cleaned by power tools. CPSES Procedure QI-QP-11.4-5, Rev. 27, paragraph 3.2.2.3, states: " Surfaces that have been power-tooled with '3-M Clean-N-Strip,' 80 grit or coarser ' flapper wheels,' sanding discs, ' roto peens,' or equivalent provide acceptable surface profile." It is alleged that:
- a. The coating systems applied to surfaces prepared using the tools specified above are not DBA qualified.
- b. These surface preparation methods provide a smoothing or polishing action, rather than a penetrating action, as obtained with sand-blasting or with a needle gun.
- c. The profile that is obtained using these methods occurs in a sparse pattern and not a densely packed pattern.
The TRT reviewed CPSES procedure QI-QP-11.4-5, Rev. 27, and subsequent revisions, ANSI N101.2, and correspondence from TUEC to the TRT, including attached copies.of related correspondence with coating manufacturers, and test reports. (Further discussion of the adequacy of procedures and methods for repair work surface preparation is provided in Coatings Cate-gory 4.) The TRT noted that ANSI N101.2 does allow surface preparation other than abrasive blasting to be used for repair work and provides test-ing methods for qualifying repair coating systems using alternate surface preparation methods. The TRT, after reviewing the supporting data provided by TUEC, observed that there was considerable evidence that power tool cleaning as a surface preparation method for repair work can provide acceptable results, although there was not complete independent laboratory DBA test data for all methods allowed and used at CPSES. Power tool cleaning does not, generally, provide the same degree and type of surface cleanliness and roughness as abrasive blasting, and if improper tools or improper methods are employed, power tool cleaning can result in a smoothed or polished surface with inadequate roughness to assure satis-factory coating adhesion. The information provided to the TRT by TUEC indicated that the most satisfactory results are achieved with 3M
" Clean-N-Strip" or 60 grit or coarser sanding devices; experience at other nuclear facilities and the industry literature support this conclu-sion. Thus, it is crucial that power tool cleaning be performed only with the proper tools used correctly and that the work be carefully inspected to assure adequate roughness and cleanliness.
Comanche Peak SSER 9 M-59
.. . _ = _ - -. --
The TRT noted that the wording in QI-QP-11.4-5, Rev. 27, permitted the use of improper tools (e.g. , "80 grit" and "or equivalent"). The TRT also noted that the inspection records for a large quantity of the repair work performed at CPSES did not provide adequate information to identify the precise tools or methods used, or to demonstrate that adequate surface roughness was achieved. (The TRT also noted that current procedures and inspection methods have remedied this deficiency for current and future . repair work.) The TRT concludes that, although power tool cleaning can result in accept-aDle Coating repair work, there is inadequate evidence to demonstrate that this type of surface preparation for coating repairs at CPSES was acceptable. However, these methods were used only for repair work in limited, but unquantified, areas. Primer Thickness of 0.5 Mils (AQ0-11). This allegation concerns DBA qualification for primer applied at a thickness of 0.5 mils. It is alleged that CPSES DCA 18,489 permits 0.5 mils thickness of primer and that this system is not DBA qualified. The TRT reviewed DCA 18,489, Revisions 0 and 1, TUEC's correspondence to the TRT on this subject, and CPSES CEL entries Nos. 8 through 15. The TRT noted that Rev. O of DCA 18,489 accepted areas of primer at a thick-ness of 0.5 mils and provided no explanation or justification for this acceptance (such.as reference to DBA qualification test data). This DCA was then revised. The TRT observed that DCA 18,489, Revision 1, does not accept or allow 0.5-mil thickness, recognizes that this thickness is un-acceptable, and dispositions the affected areas where this thickness was observed by placing them on the CEL. (Further discussion of the use of a DCA, rather than a nonconformance report, for this purpose is provided in Coatings Category 5.) The TRT reviewed the CEL, where entries No. 8 through No.18 record approximately 100 square feet of surface area exempted because primer thickness was below the specified minimum. (Further discussion of the CEL is provided in Coatings Category 6.) The TRT concluded that the allegation was correct in that a primer thick-ness of 0.5 mils initially was allowed at CPSES, and there was no evidence that primer at this thickness was DBA qualified. However, with issuance of Revision 1 of this DCA,. areas with this thickness were exempted and.not accepted. Applying Imperial System 115/1201/11S/1201 to a Thickness of 102 Mils (AQ0-12). This allegation concerns the concrete coatings repair / overlap system discussed in the evaluation of for AQ0-01. It is-alleged that, if the maximum limits given in CPSES procedure CCP-40, Rev. 5, paragraph 4.3.1.2, are used, a 102-mil thickness of the Imperial System 11S/1201/115/1201 is allowed, and that this system is not OBA qualified. The TRT reviewed all revisions of CPSES procedure CCP-40, correspondence from TUEC to the TRT on this subject, and related test reports. The TRT recognized that the total area involved could not be quantified and was not entered in the CEL, but based on the review of this information and the investigation of allegation AQ0-01, including TRT examinations of coated Comanche Peak SSER 9 M-60
concrete surfaces, the TRT found that this system was used only in limited areas, if at all. The TRT found that the test data presented provide some confidence that this system, at this thickness, will perform acceptably, but do not constitute DBA qualification for the system permitted by the procedure. The TRT concluded that while the allegation was substantiated, because of the limited area involved, this item is without safety consequences. Reactor Core Cavity _ Coatings __(AQO-13). This allegation concerns DBA qualification of coatings applied to the CPSES reactor core cavity. It is alleged that these coatings have not been DBA tested for the higher radiation exposure levels in this area, and that, in the event of a LOCA, failed coatings would interfere with the engineered safeguards systems. The TRT reviewed correspondence from TUEC to the TRT on this subject, including correspondence between TUEC and Gibbs & Hill, entry 41 in the CEL, and CPSES NCR C-83-00461, Revision 0, February 11, 1983, and Revi-sion 1, February 15, 1983 (with related March 10, 1983, telex from Gibbs
& Hill to TUEC), and also physically examined the reactor core cavity.
NCR-C-83-00461 documented the fact that reactor core cavity coatings had not been DBA qualification tested to demonstrate their ability to withstand the levels of gamma and neutron exposure which will be present in the reac-tor core cavity. The TRT noted that this NCR has been dispositioned "use-as-is" based upon the reasons provided'in the Gibbs & Hills March 10, 1983, Telex. The most significant of these reasons is that failure of coatings in this area will not interfere with the operation of engineered safeguard systems because debris from failed coatings cannot travel from the reactor core cavity, and therefore, qualification for coatings in.this area is not needed. The TRT concurred with this reasoning and confirmed it by an examination of the reactor core cavity. Because of the physical configura-tion of the reactor core cavity and its location in the plant, failed and
. flaked-off coatings would not flow out of this area during a LOCA. The
_ reactor core cavity is located below all other areas of the Containment Building and below the emergency core cooling system sump screens; the only openings to the cavity are in the ceiling. Because of the higher radiation exposure levels in this area, TUEC has assumed that the coatings will fail, and accordingly, placed this area on the CEL. The TRT. reviewed the CEL and confirmed that entry 41 documents 3135 square feet of coatings in this area. ' The allegation that reactor core cavities have not been OBA qualified for the environment in that area is substantiated. However, for the reasons given above, the TRT concurred with TUEC's analysis and disposition of this item. i Applying CZ-11 or 191P over_P_-305, and P_-305 over 1201 (AQ0-15). This allegation concerns DBA qualifications of nonstandard systems used in i repairs where coatings overlap. It is alleged that CPSES' procedure CCP-30, Revision 11, paragraph 4.4.3.0, allows CZ-11 or 191P to be applied over P-305 without sanding back to a " mottled" transition, that this same para-graph allows applying P-305-over 1201 and vice versa, and that-these i systems are not DBA qualified. i l Comanche Peak SSER 9 M-61 I l i l
The TRT reviewed all revisions of CCP-30, correspondence from TUEC to the TRT on this subject, and CPSES requests for information or clarification (RFICs) which address this subject. The TRT found that the procedures included incorrect and confusing instructions in regard to repairs where coatings overlap and that the RFICs did not clarify the situation. (Further discussion of the adequacy of CPSES procedures is provided in , Coatings Category 4.) TUEC has stated that the overlapping of coating systems, although the resultant systems have not been specifically DBA qualified, is standard practice in the nuclear industry. The TRT concurs with this statement, provided that such areas are limited to the smallest practical size; that such overlaps are performed in accordance with " standard good coating practices" and the manufacturers' instructions and are carefully inspected; and, that there is adequate technical justification for the overlapping systems and the application methods employed. The TRT evaluated overlapping systems used at CPSES and determined that their limited use was not an item of concern. However, the TRT determined that applying inorganic zinc over an epoxy coating is not technically acceptable and that there is considerable evidence in the industry that this system will .not perform satisfactorily. -TUEC estimated that the surface area coated with inorganic' zinc on top of epoxy ranges from 2,500 to 6,500 square feet; the TRT concurs with this estimate. The TRT concludes that this allegation is substantiated. However, because of the relatively minor surface areas involved in these overlap systems, the TRT concluded that this item is without safety consequences'. (The adequacy of design evaluations and justifications for coatings systems at CPSES is discussed elsewhere in this report.)
- 5. Conclusion and Staff Positions: Based on the TRT's evaluation, these allegations were largely substantiated. However, the specific coating systems described in these allegations represent a small percentage of the total area coated within the CPSES Containment Buildings. (The significance of the size of this area is discussed in Coatings Category 6.)
The TRT substantiated AQO-01 that adequate testing had not been performed to demonstrate the DBA qualification of the materials applied .in the sequence specified in B&R Procedures.CCP-40. However, because of the-limited area involved, the TRT concludes that the use of these systems was acceptable and would have no adverse safety effects. The TRT substantiated AQO-02 that nonstandard coatings sequences were not DBA qualified. -Unless noted otherwise, the TRT' finds this practice acceptable when it is limited to small areas of repair and overlap. The TRT ' substantiated AQO-03 that this. coating system is not DBA qualified. However, because the item was entered on the Coatings Exempt Log, the TRT concludes that use of this system is without adverse consequences or safety
. significance. i l
Comanche Peak SSER 9 M-62 ' 1
The TRT partially substantiated AQ0-04 in that TUEC did not provide the TRT with adequate documentation to fully demonstrate that this system is fully qualified. However, because the area involved is limited and because the item was placed on the Coatings Exempt Log, the TRT concludes that it will have no adverse safety consequences. The TRT partially substantiated AQ0-05 that TUE did not provide test data sufficient to demonstrate that this system is DBA qualified. However, based on its assessment, the TRT concludes that this system will perform adequately in service and will have no adverse safety consequences. The TRT substantiated AQ0-06 that Imperial Nutec 115 has not been DBA qualified for application over foreign objects. However, because the areas involved were limited, TUEC's treatment of this item was acceptable. Accordingly, this issue is without safety significance. The TRT has not substantiated AQO-09. Application of inorganic zinc multi-ple coats is DBA qualified, and application in three coats is not contrary to CPSES procedures. The TRT has partially substantiated AQ0-10 in that adequate evidence was lacking to demonstrate that preparing surfaces with power tools for coating repairs was acceptable. However, because this preparation method was used only for limited areas needing repairs, the TRT concludes that this practice is without safety significance. The TRT has partially substantiated AQO-ll in that a primer thickness of 0.5 mils was permitted without evidence that primer of that thickness was DBA qualified. However, with the issuance of Revision 1 to the DCA, areas of that thickness were exempted. Accordingly, the TRT concludes that this issue is without safety significance. The TRT has substantiated AQ0-12 in that the test data do not constitute DBA qualification for the system permitted by the procedure. However, because of the limited area involved, the TRT concludes that this item is without safety significance. The TRT has substantiated AQ0-13 that coatings in th'e reactor core cavities have not been DBA qualified. However, because this item was placed on the Coatings Exempt Log, the TRT concludes that this item is without safety significance. The TRT has substantiated AQ0-15. However, because of the relatively minor surface areas involved, the TRT concludes that this item is without safety significance. Based upon the TRT's broader investigation of TUEC's design and engineering for the major coating systems at CPSES, as well as the investigations of specific allegations, the TRT concludes that the performance of design and engineering functions for coatings was inadequate. Inadequate engineering review and evaluation was evidenced by the deficiencies in the DBA data base and in design control activities and documentation, and can be gen-erally characterized as a failure to make accurate technical assessments Comanche Peak SSER 9 M-63
and valid engineering judgments in regards to coatings. The TRT concludes that this engineering performance clearly demonstrates that some personnel assigned by TUEC to perform coatings engineering functions either were not capable 9r not qualified to perform these functions. The result of these deficiencies is that TUEC has not demonstrated that the coating systems applied at CPSES are DBA qualified, in accordance with ANSI N101,2 and N512. These deficiencies have also resulted in deficient coating work that could have been prevcated by proper design and effective
. engineering evaluation.
The generic implication of these conclusions is that those responsible for TUEC's coating quality assurance program failed to implement effective design control of safety-related coatings and to detect and remedy deficient. implementation; thus rendering the uverali ef fectiveness of the coatings QA program and of basic engineering functions indeterminate. These deficiencies, although not of safety significance in the coatings ! area must be considered in evaluating the effectiveness of TUEC's overall ) QA/QC program. Following completion of its onsite work, the TRT Coatings Group attempted ' to contact all of the allegers to discuss its findings of their original I concerns and to obtain any additional comments from them. A summary of I the followup interviews is included in Section 2.2.3 of this Appendix.
- 6. Action Required: None.
Comanche Peak SSER 9 M-64 l l___________
- 1. Allegation Category: Coatings 3, Traceability
- 2. Allegation Numbers: Parts of AQ0-36, AQO-34, and AQ0-62.
- 3. Characterization: It is alleged that:
- The traceability of coating materials was not always maintained (AQ0-36).
- ihe requirements of American National Standards Institute (ANSI)
Standard N45.2.2-1370 were not mat for coating material storage (AQ0-34).
- Some paint used at Comanche Peak Steam Electric Station (CPSES) in Service Level I areas was contaminated with grease and oil (AQO-62).
- 4. Assessment of Safety Significance:
Traceability Not Always Maintained (AQ0-36). Because of the general nature of allegation AQ0-36, the NRC Technical Review Team (TRT) evaluated the CPSES system for assuring coating materials traceability and the methods used for storing and handling coating materials to determine whether the system and methods could allow traceability to be lost. The TRT reviewed CPSES nonconformance reports (NCRs) pertaining to coatings, and selected four which bear on the subject of traceability; three of those NCRs are discussed under the review of AQO-62, below. The fourth NCR, C-83-02938, October 27, 1983, directly concerns loss of traceability due to the transfer of coating materials from a mixing area to application locations, without maintaining identification of the material during transport. The TRT's review of this NCR and of associated inspection reports revealed that the nonconforming condition involved small quantities of msterial that were mixed at one time then divided into different containers to be used for touch up and other coating work on a number of small areas. The NCR disposition section described a change to the method for mixing and distribution to require identification-tags to be placed on each container when material was divided after mixing. The TRT noted that this NCR concerned a relatively isolated incident which involved 12 inspection reports over a 3-day period, October 19-21, 1983, and that the coatings in question were used only for a few small areas. The TRT did not find any other examples of NCRs concerning loss of trace-ability of coating materials due to this same cause and, because measures were instituted to prevent recurrence, the TRT concluded that this parti-cular incident did not indicate any more general loss of traceability. A followup interview was conducted with alleger A-11 on February 5, 1985. The alleger indicated that the concern with disposition of NCR C-83-02938 (Allegation AQ0-36) would be better characterized as a generic concern with traceability of coatings, rather than the specific NCR disposition. Because, as noted below, the TRT conducted a generic evaluation of coatings traceability, the TRT believes the concerns expressed by the alleger have been adequately considered. Comanche Peak SSER 9 M-65
. _ . , .-- _. _ _ _ - =- _. Thethis on TRTsubject. reviewed Texas Utility Electric Company's (TVEC's) correspondence TUEC letter TXX-4201, June 22, 1984, under the heading
" Allegation No. 36. . Evaluation of Validity" states in part, " Traceability does not exist in all cases for coatings applied prior to November 1981.
These coatings, however, are within the scope of the backfit test programs which determines their adequacy." The TRT reviewed NRC Region IV Inspec-tion Report 59-445/81-15; 50-446/81-15 (for inspections performed October 13-16 and 19-23, 1981) and related correspondence and Texas Utilities Generating Company (TUGCO) QA Audit Report TCP-24, October 2, 1981, which is referenced therein. These documents reported deficiencies in records of coating work performed at CPSES prior to November 1981. The backfit test pregram was established by TUEC for the purpose of determining the adequacy of applied coatings for which existing records were deficient. The TRT observed that backfit test program inspection reports did not record coating material batch numbers or reference any other records where this information was recorded. Further discussion and assessment of the CPSES backfit test program is provided in Coatings Category 1. The TRT determined that the backfit test program and the backfit inspec-tion records do not restore traceability of applied coatings for which deficient previous records cannot provide traceability. The backfit test program inspections determined the adhesion and thickness of the applied coatings at the time those inspections were performed. These two physical attributes do not demonstrate that the applied coating materials are certified nuclear grade materials which have been manufactured, stored, mixed, and applied without alteration or contamination, and do not demonstrate that these materials are qualified. The TRT evaluated the coatings material control and traceability system to determine whether it ensured that coating materials applied at CPSES were traceable to batches which had passed qualification tests. An adequate traceability system must involve: (1) manufacture of the material under a manufacturer's Quality Assurance Program; (2) certification by the manufac-turer that the batch manufactured is essentially identical to the batch which was qualification tested; (3) receipt, control, storage, and use of the material by TUEC under a system that assures that the materials are handled and used, as recommended by the manufacturer, in a manner essen-tially identical to the manner in which the originally tested batch was used; (4) implementation of an inspection and documentation system which provides records to demonstrate all of the above. The TRT identified the following procedures which include instructions for handling, storage, use, and control of coating materials: TUGC0 procedure CP-QP-8.0 and Brown & Root procedure CP-QP-8.1 provide instructions for receipt inspections at CPSES for all incoming items, including coating materials. Brown & Root procedure CP-CPM 8.1 provides instructions ing materials.for receipt, storage, and issuance of items, including coat-Brown & Root procedures CCP-30, CCP-30A, and CCP-40, which provide overall instructions for performing coating work at CPSES, include specific instructions for storage and dispensing of coating materials and provide storage temperature limitations. TUGC0 Procedure QI-QP-11.4-17 provides instructions for monthly surveillance by QC of coating material Comanche Peak SSER 9 M-66
storage. TUGC0 Procedure QI-QP-11.4-22 provides spe'cific quality control measures which govern the transfer of protective coatings TUGC0 identification procedures QI-QP-11.4-1, numbers when cutting coated stock materials. QI-QP-11.4-5, and QI-QP-11.4-10, which provide overall i materials just prior to mixing, and during mixing and application. The TRT observed that, although most of these procedures had undergone numerous revisions during the construction phase at CPSES, the overall system for coating material storage, handling, and control had remained substantially the same throughout. The normal site routine for handling coating materials, which is described by the procedures and which was observed by the TRT, is a site warehouse area by site receiving QC inspectors in accordance with standard site receiving routines that apply to all incoming safety-related items. The materials are then issued to the coating storage area near the paint shop. The coating storage area has separate storage for "Q" and "Non-Q" paint materials, is temperature controlled, and is equipped with recording thermometers. The coating storage area and storage functions are operated by coatings craft personnel (not QC), and are subject toWhen coating m monthly surveillance by coatings QC inspectors. are needed, they are either mixed in the coating storage area, then trans-ported to the site where they willAbe applied; coatings QCor, transported inspector tothe verifies a mixing area near or at the use location. acceptability of the material immediately prior to mixing and observes the mixing and application process to verify conformance with requirements. In those cases where mixing is not performed at the application site, the QC inspector who inspects the mixing process attaches a mixing record to the material container to transfer mixing inspection information to the inspector at the application site. To assess TUEC's implementation of this system, the TRT randomly selected approximately through 1984. 20 CPSES coating work inspection reports writte were recorded on these inspection reports, although in some cases there were other deficiencies in these inspection reports which rendered trace-ability indeterminate. For example, Inspection Report PC01279 documents the application of coatings to an area of the Unit 1 Reactor Building 27, 1978, and has an attached mixing (presumably liner plate) on Junerecord (also identified as PC01279), which materials for this area on September 8, 1978. A similar situation exists in Inspection Report PC03650, which documents surface preparation and primer application on December 1, 1977, of coating materials which w mixed on November 22, 1977. it was physically impossible for these materials to have been applied as recorded in the inspection reports due to the time discrepancies between mixing and application. (Materials cannot be' applied before they are mixed, and can only be applied during a limited " pot life" The TRT also observed many other deficiencies in the used at CPSES.) inspection reports, including inadequate description or location of areas M-67 Comanche Peak SSER 9
or items coated, improper changes and corrections, lack of signatures or acceptance initials at interim and final inspection points, and lack of date/ time correlation between various sections of the reports. In many of these cases, the TRT observed, the deficiencies noted were serious enough to render the inspection reports unacceptable as quality records and inade-quate to provide documentation of material traceability. The TRT noted that some deficiencies in coatings records for the period prior to November 1981, were identified previously in NRC Region IV Inspection Report 50-445/ 81-15; 50-446/81-15 and TUGC0 QA Audit Report TCP-24, October 2, 1981. The TRT also noted that traceability deficiencies for that period have been acknowledged by TUEC in correspondence to the TRT. The TRT next selected approximately 30 batch numbers of coating materials taken from the inspection reports just discussed and from certain noncon-formance reports. The TRT traced these batches back to receiving inspec-tion reports (RIRs) which document their initial receipt at CPSES. The TRT noted during this process that TUEC had an efficient computerized tracking system to allow identification and retrieval of both coating work inspection reports and receiving inspection reports by coating material batch numbers. The TRT found that, in general, the receiving records were in order, and the receiving records packages contained all pertinent infor-mation, including copies of coating manufacturers' product identity certi-fication records, as required by ANSI N101.4 and ANSI N101.2. However, the TRT could not locate an RIR or manufacturer's certification for batch number 0M1708M of Carboline Carbo Zinc 11 (which was identified in NCR C-81-01724) . The TRT also examined the receiving records for the first shipments of coating materials from Carboline, Ameron, and Imperial, the manufacturers of Service Level I steel and concrete coating materials. These RIRs and attachments included evidence of receipt inspections and checklists, manufacturers' certifications, invoices, evidence.of source surveillance by Brown & Root or waiver of same, initial storage location at CPSES, and notation of the governing TUEC purchase order. The TRT next reviewed vendor files for Ameron, Carboline, and Imperial. These files included evidence of initial evaluation and approval of these vendors by Brown & Root based upon QA-Department Survey Checklists comple-ted by the vendors, review and approval of the vendors' QA Program Manuals by Brown & Root, and facilities surveys by Brown & Root. These files also included records of regular audits of these manufacturers by Brown & Root from 1977 through 1981. Although these audits occasionally included find-ings of deficient conditions, documentation was included which demonstrated corrective actions and verification of acceptability via re-audit. The TRT found that since 1981, regular audits have been performed by TUEC, rather than by Brown & Root. The vendor files also contained copies of the original purchase orders to these vendors together with subsequent changes. The TRT found that these purchase orders were generally complete and acceptable, and included the imposition upon the vendors by TUEC of appropriate regulatory requirements and industry standards (e.g., 10 CFR 50 Appendix B, ANSI N45.2, ANSI N101.4, ANSI N101.2, and 10 CFR 21). The vendor files also included general cor- .' respondence sections which showed evidence that there had been vendor participation in a pre-job meeting, as required by ANSI N101.4, and in one Comanche Peak SSER 9 M-68 1 r, , - ~ . ~ , , ,*- . r - - - ,
case showed evidence that the manufacturerHowever, had transmitted reports of in-the TRT observed house laboratory qualification tests to TUEC. that these files did not include complete qualification testing data nor(This subject evidence of a review and evaluation of that data by TUEC.is discuss described The TRT found that the system and its implementation by TUEC, as above, included most of the important measures required to control mate-rials and to maintain traceability. However, as noted below, the TRT observed certain deficiencies in the procedurally described system and in practices which are not procedurally addressed that could allow traceabil-ity to be lost. Coating storage areas are under the management of craft production person- ; nel (access to stored materials is not controlled independently by QC personnel), and "Q" and "non-Q" coating The result of this materials are identified arrangement only by is that once their separate storage areas.a container of unmixed material is loaded onto a truck for tran where, it is no longer possible to determine whether the material is "Q" or "Non-Q." An inspection of the containers, performed immediately prior to mixing, would not detect such a mixup in those cases where there is no evident difference in the container labels, such as "Q" materials which had earlier been moved to the "Non-Q" storage area due to athat (Note problem no with certification or because of a previous loss of control. In a similar tagging or label change is required when this occurs.) manner, there is no independent control by QC personnel to assure that unacceptable materials returned to storage from another location are, in fact, returned to "Non-Q," rather than "Q" storage, or to prevent the l transfer of materials between adjacent "Q" and "Non-Q" storage areas. The storage area temperature record charts which are attached to the monthly storage QC inspection reports are not identified as to re the "Q" or "Non-Q" storage area, or some other location). Some paint mixing slips, describedThus, above, are not retained as the original area transcribes the information to his own report. record of the mixing inspection, including the inspector's signature, is lost. There is no clearly assigned responsibility for maintaining inventory and storage records for coating materials after transfer from the site ware-house area to the coating storage area, and the monthly QC inspectionThus, the official reports do not identify what materials are in storage. records do not sh the storage area during a given time period, nor is there any any specific batch of material. There is no procedurally required control of coatings materials which have been moved from the coating storage area to various application or mixing areas in the plant. Thus, coatings may in fact be kept in such locations M-69 Comanche Peak SSER 9
for extended periods of time without temperature cont ol or records or QC surveillance. The TRT did not find any specific evidence the these deficiencies re-sulted in any general loss of material cont si or traceability (after 1981), except as specifically indicate aerein. However, it is evident based on this review that a more "igorous system of coatings material control is necessary to prov % confidence that materials do not lose traceability throuch reafusion, alteration, loss of status control, or exposure to environmental temperatures outside of the specified. range, during storage and handling on site. Requirements of ANSI N45.2.2 Not Met (AQ0-34). Allegation AQ0-34 concerns control of the storage and handling of coating materials. It is alleged that the requirements of ANSI N45.2.2 were not met for the storage and handling of coating materials. In assessing this allegation, the TRT reviewed the CPSES Final Safety Analysis Report (FSAR), ANSI N45.2.2, and correspondence from TUEC to the TRT on this subject. The CPSES FSAR, Section lA (B)-16, specifically states.that the require-ments of NRC Regulatory Guide 1.38, and thus, ANSI N45.2.2, will not apply during the construction phase. The TRT noted that alternate methods for controlling the storage of coating materials are provided through CPSES FSAR commitments to compliance with ANSI N101.4'and related standards, and are detailed in CPSES Specification 2323-AS-31 and pertinent CPSES proce-dures. The adequacy of those alternate procedural requirements and their implementation is discussed in the assessment of AQO-36, above. Application'of Contaminated Paint (AQ0-62). Allegation AQ0-62 concerns alteration or contamination of coating materials prior to application. Such alteration or contamination renders a coating material unqualified and untraceable in that it is no longer essentially identical to the batch which was subject to original qualification testing. It is alleged that some paint used at CPSES in Service Level I areas was contaminated with grease or oil, prior to application, and was applied anyway. The TRT-found and examined three separate' instances of possible contamina-tion of coating materials. All three occurrences were documented in non-conformance reports. NCR C-1729, October.2, 1979, concerned the,following nonconforming condition: " Carbo-Zinc 11- base-batch 9H5381M was noted for containing what is believed to be grease particles and/or foreign contami-nants. Discrepant Qty - 300 gal /5 containers." The-dispositon of the original version of this NCR was "use as is" (after_ straining to remove all contaminants). This disposition was based upon a Telex'from the manufacturer, Carboline, which was attached to the NCR. Revision 1 of this NCR, issued on October 17, 1979, changed the disposition to " Return to Vendor," due to the fact that the straining process had been unsuccess-ful in removing the contaminants. To determine whether any of this contaminated material had been " applied anyway," as alleged, particularly in the time frame between the original ~ issue of the NCR and Revision 1, the TRT examined coating inspection records for work done during September, Comanche Peak SSER 9 M-70
October, and November 1979. The TRT found no evidence that the contami-nated material, batch 9H5381M, was ever applied. The TRT noted that the Telex from Carboline, which was provided as justification for the strain-and-use-as-is disposition, did not provide any technical explanation by Carboline to indicate why Carboline felt that the presence of the particles would have no adverse affect upon the material, nor any indication that Carboline had reviewed the NCR or was aware that the particles had been identified as " believed to be ' grease'." NCR C-81-01724, December 14, 1981, and Revision 1 of the same NCR, also dated December 14, 1981, document indeterminate conditions in batches OM2708M, OL2531M, and 1J2791M of Phenoline 305 Catalyst, Part B. Batch OM2708M was identified as being " dark wine color," rather than the normal amber color, and all three batches had unknown particles in the material. This nonconformance report was dispositioned "use as is" based upon communications from Carboline, which were attached to the NCR. The TRT reviewed the information attached to this NCR and found a situation similar to that discussed above for NCR C-1729, in that the Carboline statements did not provide adequate justification or assurance that the identified deficiencies would not have adverse consequences. Carboline's statements did not discuss the nature of the particles, but indicated that these particles probably got into the coating material in their plant. In a Telex dated December 31, 1981, Carboline recommended that the batches not be used until they had determined ^the cause of the color difference. This recommendation was not included in their letter of January 19, 1982, although there was no indication that the cause had been determined. NCR C-81-01673, of December 6, 1981, documents a condition observed during mixing of Carboine Phenoline 305, base batch 1J2789M, curing agent batch 1F1054M, and thinner batch 1E1861M. The NCR reports that after repeated hand and power agitation the materials would not thoroughly blend, and dark green pigment floated to the top. The NCR also notes that these materials had already been used to coat nine electrical supports. This NCR was dispositioned "use-as-is," based upon a visual examination of the nine electrical supports which revealed no color abnormalities, and on a statement by the engineer who wrote the disposition that such " pigment float" sometimes occurs but will not affect the cure or integrity of the coating. The disposition further requires that an agitated pot shall be used during application if pigment float is noted. Upon reviewing this NCR and its attachments, the TRT found no clear ex-planation of the cause of the observed condition, other than the engineer's statement that pigment float sometimes occurs, especially when higher levels of thinner are used, nor any justification for the statement that this condition will not affect coating cure or integrity. The TRT noted that the inspection of the electrical supports to which the coating had been applied provided some confidence that the cure and integrity were not affected. Nonetheless, it is the TRT's opinion that a more rigorous investigation of possible contamination or alteration, and communication with the manufacturer on this subject, would have been appropriate prior to the issuance of a use-as-is disposition. Comanche Peak SSER 9 M-71
. - __ . __ . , _ _ _ _ __ ~ ._ ~ ._ _ . _ . ._ , _
Nuclear-certified coating materials are expected to appear and behave uniformly without variation from batch to batch or container to container; any observed differences or irregularities, such as those documented in these NCRs, should be thoroughly investigated to assure that traceability to the original qualification tested batch has not been lost through deficiencies in manufacturing, handling, or use.
- 5. Conclusions and Staff Positions: Based upon the reviews of individual allegations described above, the TRT concludes that:
Allegation AQ0-36 that the traceability of coatings materials was not always maintained is substantiated. The primary significance of this allegation is that, at present, coatings applied in areas included in the backfit test program are of indeterminate traceability. Allegation AQ0-34 concerning noncompliance with ANSI N45.2.2 is substan-tiated. However, compliance with that standard is not required by the CPSES FSAR. Therefore, this allegation, as stated, has neither safety significance nor generic implications. The TRT noted that alternate methods for controlling storage of coating materials were provided by TUEC procedures in compliance with ANSI N101.4. The adequacy of those proce-dures and their implementation is discussed above in the assessment of AQ0-36. Allegation AQ0-62 that coating-materials which were contaminated with grease were applied at CPSES was not supported by the TRT's investigation. In the one instance where " grease" contamination was observed,.the mate-rial was returned to the vendor and the TRT found no evidence to indicate that any of this contaminated material was applied. Therefore, although partially substantiated, this allegation has neither safety significance nor generic implications. However, in the investigation of the subject of contaminated materials in general, the TRT found instances of coatings which had been applied with unknown contaminants or which exhibited other unusual and unexplained properties. The TRT found that in these other cases, the operation and documentation of the nonconformance reporting system at CPSES was inade-quate to demonstrate the ccuse of the irregularities,.to provide adequate technical justification for use-as-is dispositions for the deficient materials, or to address and correct whatever fault in the manufacturing, handling, or materials control processes produced the deficient materials. All three allegations assessed raise questions about TUEC's material control and traceability system. Based on its review of that system, the TRT concludes that the material control system, as it is implemented cur-rently at CPSES, provides general control and assurance of traceability for most coating materials, but is not sufficiently rigorous to assure traceability for all coating materials. The primary area of concern, as discussed above, is that materials storage and issuance routines do not include the necessary controls to assure segregation of "Q" and "Non-Q" materials and to prevent the use in Service Level I areas of materials designated "Non-Q." The TRT was not able to identify specific incidents wherein this lack of rigorous control had resulted in misuse of materials; Comanche Peak SSER 9 M however, procedural changes would be needed to prevent such incidents in the future. (The subject of procedures is discussed further in Coatings Category 4.) The TRT determined that the material control system was not adequately implemented prior to November 1981. NRC Region IV Inspection Report 50-445/81-15; 50-446/81-15 has identified violations concerning inadequate inspection and documentation practices for coatings work prior to November 1981, which resulted in loss of material traceability. The TRT has deter-mined that the corrective action performed by TUEC to date (i.e., the backfit test program) has not restored material traceability. These deficiencies, although not of safety significance in the coatings area, must be considered in evaluating the effectiveness of TUEC's overall QA/QC program. Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original concerns and to obtain any additional comments from them. A summary of the followup interviews is included in Section 2.2.3 of this Appendix.
- 6. Actions Required: CPSES nonconformance reports (NCRs) C-81-01724 and C-81-01673 provide "use-as-is" dispositions for discrepant coating mate-rials, with inadequate technical justification for the disposition.
Accordingly, TUEC shall provide adequate technical justification to demon-strate the acceptability of the batches of coating materials listed in these NCRs or, alternatively, TUEC shall identify and quantify the areas where these batches were used and place these areas in the coatings exempt log. Additionally, TUEC shall review all other NCRs which concern discre-pant or irregular conditions in coating materials. For any such NCRs which were dispositioned "use-as-is," TUEC shall identify the batches and provide adequate technical justification for their acceptance, or identify and quantify the areas where the batches were used and place these areas on the coatings exempt log. The coatings exempt-log will be used in planning future inspections of coatings consistent with the guidelines in Appendix L. Comanche Peak SSER 9 M-73
. - . . . ._= - ._ - - _.
Allegation Category: Coatings 4, Coatings Procedures 1. Parts of AQO-01, AQ0-02, AQ0-03, AQ0-04, AQ0-05,
- 2. Allegation Number:
AQ0-06, AQO-07, AQ0-08, AQ0-09, AQ0-10, AQ0-11, AQ0-12, AQ0-15, AQ AQ0-18, AQ0-19, AQ0-20, AQ0-23, AQ0-24, AQ0-3 AQ0-51, AQ0-58, and AQ0-62. Characterization: It is alleged that: 3. a. Comanche Peak Steam Electric Station (CPSES) coating work procedures and instructions allowed the use of coating systems and Theapplication methods that were not Design Basis Accident (08A) qualified. instances identified in the allegations are: Procedure CCP-40, paragraph 4.3.1.2, allows application of 115/1201/115/1201 or Imperial coatings in the sequential order i 11S/1201/11/1201 (AQ0-01). and Table A2 1 Nonconformance Report (NCR) C83-01752, June 23,1983, in Appendix A of CPSES specification 2323-AS-31, allow repair coatings to be applied in sequences different from the original application sequences (AQ0-02).
- Design Change Authorization (DCA) 17,142, Revision 2, allows application of Carboline Phenoline 305 (P-305) over another manuf acturer's epoxy coating (AQ0-03).
DCA 12,374, Revision 1, allows Carboline Carbo Zinc 11 (CZ-11) to be topcoated with Imperial 1201 (AQ0-04). Procedure CCP-30A, Revision 2, paragraph 1.3.1, allows applica-tion of P-305 over Ameron Dimetcote 6 (D-6) (AQ0-05). Procedure CCP-40, Revision 5, paragraph 4.1.1. in concrete (AQ0-06). Inorganic zine application in three coats is allowed (AQO-09). DCA 18,489 allows primer to be applied to a thickness of 0.5 mils (AQ0-11). Procedure CCP-40, Revision 5, paragraph 4.3.1.2, allows the Imperial coating system 11S/1201/11S/1201 to be applied to a thickness of 102 mils (AQ0-12). Procedure CCP-30, Revision 11, paragraph 4.4.3.0, allows CZ-11 or Carboline 191 primer (191P) to be applied over P-305, and P-305 to be applied over Imperial 1201 (AQ0-15). b. Coating work procedures and related Specific documents allegations are: include instruction which are technically incorrect. M-75 Comanche Peak SSER 9
NCR C83-01986 provides a repair disposition for cracking and flaking of concrete coatings which will not remedy the cause of the deficiencies (AQ0-07).
- Procedure CCP-30, Revision 11, allows inorganic zinc primer to be applied over zine residue, a practice which will cause i adhesion problems and prevent galvanic action (AQO-08).
TUGC0 procedure QI-QP-11.4-5, Revision 27, paragraph 3.2.2.3, permits acceptance of nonqualified and technically inadequate power tool cleaning methods for surface preparation (AQ0-10). The method used to remedy high dry film thickness (DFT) of CZ-11 will result in poor adhesion of the topcoat (AQ0-38). Applied P-305,1 and 2 years old, was topcoated with new P-305 with little or no surface preparation (AQO-39). Residues resulting from power tool cleaning of surfaces were removed by improper methods (AQ0-40). A foreign cleaning solution was used to wipe surfaces immediately prior to repairs (AQ0-41). The methods used at CPSES to verify the cure of inorganic zinc primers are not adequate, and inorganic zinc primers are not properly cured prior to topcoating (AQO-43). The " nickel test" for verifying the cure of inorganic zinc primers prior to topcoating was not performed properly (AQ0-44).
- c. Backfit inspection procedures and methods were inadequate. Specific i allegations are:
QC inspectors were not allowed to identify visual defects during backfit inspections (AQ0-18). TUGC0 procedures QI-QP-11.4-23 and QI-QP-11.4-24 are very vague regarding the way that backfit inspections are to be conducted (AQ0-19).
- Adhesion tests of protective coatings were not properly performed (AQ0-20).
Excessive thinning of P-305 resulted in a weak and brittle film and made it impossible to obtain a Tooke gauge reading (AQ0-51).
- d. Coatings were applied to surfaces where they should not have been applied. Specific allegations are:
"Q" coatings have been placed over rusty, scaly, unprepared metal surfaces inside pipe supports made of tube steel without end-caps (AQ0-24).
Comanche Peak SSER 9 M-76
Imperial 115 and 1201 were applied )ver duct tape and foam ru on Richmond Inserts (AQ0-42). Coatings were applied over joints which were filled with foam (AQ0-48). Overspray was allowed in areas which had been inspected previously (AQO-49). Specific allega-
- e. Quality Control (QC) inspections were inadequate.
tions are: Tests of the cleanliness of compressed air used for spray appli-cation of coatings were invalidated due to the practices of production personnel (AQ0-17). The coatings QC program is inferior to such programs at other nuclear power plant projects (AQO-23). QC management interpreted an SSPC-SP-6 blast requiremen on a DCA as "do the best you can" and told the QC inspectors "not to worry" about difficult access areas (AQ0-31). Repairs of defects have been accomplished with no reinspecti of the defects (AQO-45). QC inspection procedures require that inspections be perfor with inadequate light (AQO-58). f. Procedures were inadequate to assure coating materials traceabi Specific allegations are: were not met for The requirements of ANSI /ASME N45.2.2-1978 material-storage (AQ0-34). The traceability of coatings materials was not always maintained (AQ0-36). Some paint used in Service Level I areas was contaminated wi grease and oil prior to application and was applied anyway (AQ0-62). The majority of allegations in-this have
- 4. Assessment of Safety Significance:
category address both DBA qualifications andThe procedures DBA aspects which may permitted the application of non-DBA qualified coatings. of these allegations are addressed in Category 2. In assessing these allegations, the NRC Technical formanceReview Team t bility reviewed the procedures and instructions which governed the per and inspection of protective coating work to determine i of the accep a ! of those procedures and instructions and to verify theThe for determin-inclus on TRT l appropriate quantitative and qualitative accept M-77 Comanche Peak SSER 9
reviewed the measures established to control the review, approval, and issuance of instructions and procedures and their revisions. The TRT also observed coating work and reviewed applicable records to determine whether important activities had been accomplished in accordance with the govern-ing procedures and instructions. The TRT performed a detailed review of procedures related to specific allegations grouped in several generic or functional areas. This review included an evaluation of both the adequacy of the written procedures and the normal practices followed when performing coating-related activities. Many of these allegations have also been considered separately in other Coatings categories; they are included here for consideration of their procedural elements. General Review of CPSES Coating Work Procedures and Procedure Control System. To evaluate the methods used te control the review, approval, and issuance of instructions and proceoures and their revisions, the TRT interviewed responsible supervisors, engineers, and managers and reviewed correspondence and applicable procedures. The TRT found that work instruc-tions and requirements for coatings applied at CPSES were provided by Brown & Root (B&R) procedures CCP-30, CCP-30A, and CCP-40. Inspection methods and requirements were provided by a number of procedures with QI-QP-11.4 designations; specific inspection procedures and requirements are identified below. The TRT reviewed the following Texas Utilities Generating Company (TUGCO) procedures, which provide requirements for the review, approval, and issuance of procedures used for coatings inspection:
- CP-QP-3.0, Revision 5, "CPSES Site Quality Assurance / Quality Control Organization."
- CP-QP-6.0, Revisions 2, 3, and 5, " Preparation of Quality Procedures and Instructions."
- CP-QP-7.1, Revisions 2, 6, and 9, " Issuance and Control of Quality Procedures and Instructions."
The TRT reviewed the following B&R procedures, which provide requirements for the review, approval and issuance of procedures for coating work operations:
- CP-CPM-6.1, Revision 4, " Preparation & Approval of Construction Procedures & Instructions."
- DCP-3, Revision 18, "CPSES Document Control Program."
Although the TRT found that these procedures were implemented in a manner that satisfied specific requirements, they did not provide adequate controls for coating work in the following areas:
- Inspection procedures reference' construction procedures and require inspectors to verify requirements or use methods described only by the Comanche Peak SSER 9 M-78
construction procedures. However, the TRT found that these referenced construction procedures are not issued to inspection personnel as con-trolled copies. (See the discussion of allegation AQO-44.) The TRT found that the qualifications of personnel who review and approve procedures do not demonstrate their ability to perform these functions. (See Coatings Categories Nos. 2 and 7.) In the CPSES FSAR, TUEC commits to compliance with ANSI N101.4-1972. This standard requires that the coating manufacturer approve applica-tion procedures. The TRT found that such approvals are only periodi-cally obtained for construction procedures, and that such approvals are not controlled in a manner adequate to assure their proper use and retention. Further, the TRT found no evidence of manufacturer approval for important revisions to requirements. Review of Individual Allegations. a .' Allegations AQ0-Ol', AQ0-02, AQO-03, AQ0-04, AQ0-05, AQ0-05, AQ0-09, AQ0-11, AQ0-12, and AQ0-15 are all concerned wi*.h Design Basis Acci-dent (DBA) qualifications of coating materials, systems, and proce-dures. The allegations, in general, are that the governing procedures and instructions allowed the use of coating systems and. application methods that were not DBA qualified. Most of the allegations in this
. group concern nonstandard coating s'ystems used for coating repairs or overlaps between different systems or between repairs and the original coatings.
A detailed evaluation of each of these allegations is provided in Coatings Category 2. The following is a summary of the TRT's findings in regards to these specific allegations, as reported in' Coatings Category 2. The TRT found that the coating systems described by the allegations in this group, except for AQ0-9, were not supported by adequate DBA testing data. However, the systems described by allegations AQ0-03, AQ0-04, and AQ0-11 were properly handled by TUEC by entering the affected items into the protective coatings exempt log (CEL).
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Consequently, the TRT found it unnecessary to review the procedural elements of allegations AQ0-03, -04, and -11. . The systems described in allegations AQ0-01, AQ0-02, '@05, AQ0-06, AQ0-12, and AQ0-15 were used only in limited'a46 e' coating overlaps and repairs. Allegation AQ0-09 is unsdn ,W . and'in that application of inorganic zinc in three coats is not a violation.of procedures and is not tech-
~
nically incorrect or contrary to data provided by DBA testing.
'The TRT's evaluation of these allegations in Category 4 focused upon the use of coating systems for which DBA qualification testing data were inadequate. The TRT was concerned with the adequacy and appro-priateness of procedures and methods which permitted the use of these i nonqualified coating systems. In investigating this subject, the TRT reviewed pertinent CPSES procedures and specifications, referenced
)
Comanche Peak SSER 9 M-79 i
below, the specific design change authorizations (DCAs) and related supplementary instruction documents described in the allegations, and the governing American National Standards Institute (ANSI) standards, ANSI N101.4, N101.2, and N512. The TRT also evaluated the subject of coating system overlaps, interfaces, and repairs, from a technical standpoint and with consideration of standard industry practices at other nuclear power plants. The governing ANSI standards provide specific methods for qualification testing and evaluation of coating systems, and ANSI N512, section 5, "Repairability and Maintenance Test," provides for testing of repair systems and methods in accord-ance with the same test methods and evaluation criteria which are used for testing the original coating systems. The TRT found that through-out ANSI N101.2 and N512 reference is made to the coating manufacturer's recommendations as a basis for procedures for coating work, including repairs, both for the preparation of test specimens and for the per-formance of actual coating work in nuclear plants. The TRT recognizes that coating manufacturers' recommended procedures for the use of their products cannot be comprehensive enough to address every possible circumstance that may occur during the use of those products. The TRT also recognizes the physical impossibility of performing complete, independent DBA testing for every possible coating method, repair system, or combination of coating materials which may occur where different coating systems interface in actual field applications. The TRT acknowledges that standard industry practice has been to accept the use of nonstandard, non-DBA-tested systems for overlaps and interfaces, while limiting such areas to the smallest practical size. Nonetheless, the TRT considers it to be important that such nonstandard coating systems be evaluated for technical viability and to assure they do not include materials, methods, or subsystems which have been shown by previous testing to be incapable of withstanding DBA conditions. It is also important that such special cases are explicitly addressed in procedures and instructions which are approved by technically qualified individuals and by the coating manufacturer, as required by ANSI N101.4. The TRT determined that TUEC's system for review of procedures by technically qualified personnel and by the coating manufacturers was inadequate, as discussed under " General Review of CPSES Coating Work Procedures and Procedure Control System" above, and-Coatings Category
- 2. The TRT also determined that TUEC's procedures and instructions in regards to special coating systems for overlaps and repairs, were complex, confusing, inappropriate, and' inadequate in many cases.
Allegations AQ0-02 and AQ0-15 are concerned with procedures for inter-faces between coating systems applied over steel where repairs are performed and where different coating systems join. The TRT found that interfaces were first procedurally addressed in B&R procedure CCP-30, Revision 9, DCN 3, November 16, 1981, paragraph 4.4.3.0. (The TRT noted that earlier instructions did include a requirement to feather the edges of repaired areas.) The TRT reviewed the evolution l of this procedural requirement through many subsequent revisions and supplementary instructions, including: a Brown & Root request for-Comanche Peak SSER 9 M-80 _ - _ _ _T
information or clarification (RFIC), unnumbered, January 7,1983; CPSES NCR C83-01752, July 5, 1983; CCP-30, Revision 11, August 16, 1983; CCP-30, Revision 11, DCN 2, October 19, 1983; RFIC, unnumbered, October 20,1983; CCP-30, revision 11, DCN 4, November 8,1983; CCP-30, revision 12, March 4, 1984 (wherein the requirements are moved to paragraph 6.1.3). The TRT observed that these changes, revisions, and clarifications involved changes in the applicability of interface instructions to different cases, changes in limitations on the width of interface areas, explanations of which coatings were permitted to be applied over which other coatings, and the like. The TRT found that many of these changes did not provide improvement or clarification, but caused confusion and raised additional questions. Numerous versions of these instructions included the statement:
"Within the interface area, overlapping of any materials or systems is acceptable." Certain unqualified combinations of coatings such as the application of one manufacturer's qualified epoxy on top of another manufacturer's qualified inorganic zinc, or the overlapping of one manufacturer's epoxy onto another's (within certain total thickness limitations), can reasonably be expected to cause no diffi-culty, based upon industry experience and inferences from related test data. On the other hand, certain combinations and sequences-of materials are technically incorrect; for example, inorganic zinc applied on top of epoxy will not exhibit adequate adhesion, and excessive thicknesses of either zinc or epoxy, which may occur at overlaps between systems or surrounding repairs, may result in crack-ing or flaking, or both. The net result of the continuously changing and confusing instructions in this area is that existing overlap areas can include every possible combination and sequence of coating materials ever used because the degree of control over application in any given overlap area is impossible to determine. The TRT deter-mined that this situation could have been prevented by clearly written instructions which delineated, at the outset, certain combinations of materials that were permissible at interfaces and certain combinations that were not, and requirements for extra care in application and inspection of interfaces to assure that these requirements were met A similar situation exists in regards to allegation AQ0-06 on the subject of coatings applied over foreign objects embedded in concrete.
The TRT recognizes that small areas of embedded " wood fuzz," metal objects, and the like are sometimes found in concrete, that procedures must address how coatings should be applied in such cases, and that it is not necessarily practical or possible to obtain complete DBA testing data for every possible circumstance. In examining applicable procedures and correspondence from TUEC in regards to this allegation, the TRT found considerable evidence that TUEC had considered the i subject and obtained manufacturer's recommendations. However, this ! information does not always appear in the appropriate procedures. TUEC's letter TXX-4201, June 22, 1984, includes a copy of an Imperial l Report 462-1-81 of Oak Ridge National Laboratories DBA testing. This l report indicates that Imperial Nutec 115 provides reasonably good performance over abrasive-blasted steel surfaces, but inadequate ) 1 Comanche Peak SSER 9 M-81 l
performance over power-tool-cleaned steel surfaces, and includes Imperial's recommendations that 115 should not be applied over power-tool-cleaned steel surfaces with suitable roughness greater than 2 square inches. However, B&R procedure CCP-40, revision 5, allows the application of 11S over embedded steel objects of up to 4 square inches which have been ground flush or smooth. Regardless of the fact that 115 over steel is not a qualified system, and regardless of the fact that the areas involved are minor, although numerous, the TRT considers that proper technical evaluation shculd have resulted in a procedural requirement to assure that 'small steel objects embedded in concrete were adequately roughened by localized abrasive
- blasting or other methods, prior to the application of 115.
- b. Allegations AQO-07, AQ0-08, AQ0-10, AQO-38, AQ0-39, AQ0-40, AQ0-41, AQ0-43, and AQ0-44 are all concerned with procedures, instructions, and related documents which provide methods and requirements for performing coating activities which are alleged to be technically incorrect. Although some of these allegations are evaluated in other coatings categories, they have been included here, as well, for the purpose of evaluating the adequacy of the related procedural requirements.
Repair of Cracked Concrete Coatings (AQO-07). This allegation identi-fies NCR C83-01986, concerning the cracking of concrete coatings, and disputes the disposition of the NCR. The disposition states, in part, that " cracking of coatings is due to excessive stresses in the coating during drying and curing," and directs repair of cracks in accordance with existing procedures. It is alleged that this repair method will not remedy the condition that caused the cracks and that the cracks may recur. The alleger has identified the underlying cause of this nonconforming condition to be foreign matter in the concrete beneath the coatings. (This general subject, embedded foreign materials, is discussed in detail under AQ0-06, above, and in coatings Category 2.) During its overall examination of coated concrete surfaces at CPSES, the TRT did
. not observe any recurrence of cracking. Foreign matter beneath the coatings is no longer visible.
The TRT also reviewed documentation showing that the cracked coat-ings in the areas described in the NCR had been repaired. The TRT examined B&R procedure CCP-40, Revision 5, which was the governing procedure in effect, and found that repair methods were described in paragraph 4.3.2.5. The TRT determined that these repair provisions are generally in accordance with the manufacturer's recommendations and generally accepted coating repair practices. The TRT determined that the documented disp'osition was an acceptable remedy for the deficiencies reported in NCR C83-01986. The TRT consitered whether TUEC's original application methods included adequate measures to prevent recurrence. -TUEC letter TXX-4201 responding to this allegation states, in part, "The preven-tative measure to preclude recurrence of this condition is the proper Comanche Peak SSER 9 M-82
I f application with emphasis on control of film thickness (see manu a turer's bulletin attached)." 19, 1983. TVEC letter TXX-4249, August 10, from Imperial of January i not necessary 1984, states that specific procedural revisions were because CCP-40 "was already in accordance" with these applicat on parameters. 19, 1983, The TRT compared the statements of Imperial's January ig i bulletin to the requirements The ofTRTCCP-40, found that Revision as controlling parameters which affect cracking.CCP-40 contains no requirements for additional curing of Imperial (a) Nutec 11 applied 10 to 20 mils thick or Nutec 115 applied in excess of 35 mils thick (as may typically occur at interior corners, bug holes, etc.); does not provide an equivalent test of hardness prior to (b) recoating thicker areas of Nutec 11 or Nutec 11S; ' (c) employs non-conservative interpolations of the manufacturer s curing schedule for Nutec 11, Nutec 115, and Reactic 1201. d The TRT requested that responsible site con received Imperial's application instructions, dated August 7, 1981. The TRT compared these instructions to B&R pro Examples are: tions and the procedure. (a) The procedure provides for concentrations of tri-sodium phos (TSP) for washing concrete surfaces which are well in exc TSP the manufacturer s recommendation, does not rinsing the TSP from the concrete. (b) The procedure does not provide for a final blowdown of the pre-pared surfaces with oil-free compressed air. (c) The procedure does not provide equivalent treatment of conc surfaces on which Nutec 10 was used as a curing membrane. (d) The procedure permits application at relative humidity levels above 85%. (e) The procedure uses non-conservative interpolations for cur times. (f) the procedure provides that up to 16 mils of Reactic 1201 ma be applied (the manufacturer's recommended maximum M-83 Comanche Peak SSER 9
(g) The procedure does not require that high film thickness, runs, and sags be abraded at least 2 mils below the specified maximum thickness. Zinc Over Zinc Residue (AQO-08). It is alleged that there will be coating adhesion problems and that the necessary galvanic action will fail to occur in areas where inorganic zinc primer has been applied over steel which ras a metallic zinc residue in the profile of the steel. Such an apalication is permitted by B&R" procedure shadows CCP-30, of tight residue Revision 11, paragi nh 4.1.3, which states: of primer which may remain in the profile of the previously prepared substrate are acceptable." The TRT reviewed test data provided by TUEC in correspondence to the TRT, and compared the procedures governing the repair of inorganic zinc coatings with methods that have been used successfully at other nuclear facilities. The TRT observed that there was adequate test data to demonstrate that inorganic zinc could be applied over itself and over steel and would exhibit adequate performance provided that proper procedures were followed both for cleaning and preparing the surface prior to application and for applying the new inorganic zinc coating. The TRT did not find any evidence that the application of inorganic zinc over tightly adhering residues of existing inorganic zinc on properly prepared steel would have any adverse impact upon adhesion or galvanic action. The TRT determined this was an accept-able, proven application technique. Coatings Applied Over Surfaces Cleaned by Power Tools (AQO-10). This a ' aTlegation concerns repair or touch up coating work in which coatings are applied over surfaces cleaned with power tools. TUGC0 instruc-tion QI-QP-11.4-5, Revision 27, paragraph 3.2.2.3 states: " Surfaces that have been power tooled with '3-M Clean-N-Strip,' 80 grit or coarser ' flapper wheels,' sanding discs, ' roto peens,' or equivalent provide acceptable surface profile." It is alleged that these methods will not result in a DBA qualified coating system but will provide a smoothed surface with inadequate surface roughness and, therefore, inadequate adhesion of subsequent coatings. (This subject is discussed in detail in Coatings Category 2.) To evaluate the adequacy of the procedural requirements affecting power tool cleaning for touch up work, the TRT reviewed relevant-CPSES procedures, manufacturers' recommendations, test data, and cor-respondence from TUEC to the TRT and between TUEC and the manufacturers. The TRT considers that power tool cleaning as an alternative surface preparation method for touch up ar.d repair work, while inferior in general to more rigorous methods, such as abrasive blasting, is normal industry practice for minor repairs and can produce acceptable results. Acceptable results are dependent upon appropriate methods and proce-dural instructions for performing and inspecting the work to assure the resulting surface is adequately cleaned and roughened to provide good adhesion. Comanche Peak SSER 9 M-84
e i In reviewing TUEC's letter.TXX-4201, test data and attached corres-pondence, the TRT noted that all available data indicated that good adhesion over power-tool-cleaned surfaces is dependent upon the use , of the proper tools, specifically 3M " Clean-N-Strip" or 60 grit or i coarser sanding devices. The TRT found that the allegation that the methods permitted by the section of QI-QP-11.4-5 quoted above, (i.e.,
"80 grit or coarser" and "or equivalent") will provide a smoothing or polishing action rather than adequate roughness for proper adhesion, is technically correct. These tools will not provide adequate ;
surface roughness, and the instructions to inspectors that surfaces prepared by such methods are acceptable were inappropriate. The TRT also noted that the instructions for performing repair work , provided in B&R procedure CCP-30, Revision 11 (which was in ef fect at the same time as QI-QP-11.4.5, Revision 27, quoted above), contained the same technically incorrect listing of tools to be used to p the work. or documentation to determine the extent of roughness or the specific Therefore, the TRT could not determine to tools which had been used. what extent, if any, power-tool-cleaned surfaces were unacceptably smooth due to the use of 80 grit sanding devices or their equivalents. However, the TRT recognizes that power tool cleaning was used only to prepare surfaces for repair so that the total involved is relatively ~ r minor. The TRT reviewed the current revisions of CCP-30 and 4 QI-QP-11.4-5 and found that this problem has been corrected through specification of the proper tools for power tool cleaning roughness by comparison against an approved visual standard. Allegation AQ0-38 is con-Grinding Excessive CZ-110FT (AQ0-38). cerned with the use of power grinding to reduce unacceptably high It is alleged i thickness of applied CZ-11 to an acceptable thickness. that this method will result in a burnished or polished surface that would cause poor adhesion of the topcoat to the zinc. The TRT reviewed relevant CPSES procedures, correspondence The from TUEC, recommendations of the manufacturer, and industry standards. accepted method for remedying high DFT of inorganic zinc is by abrad-ing, sanding, screening, or performing other appropriate mechanical TUEC's proce-methods to reduce the thickness to acceptable levels. Burnishing or polish dures specify these methods. zince resulting from these methods generally have no Adverse effects upon topcoat the topcoat to the zinc coating. adhesion will be increased, however, if the inorganic zinc has aged for several years prior to being power ground to reduce thickness. This allegation is concerned with the Recoating P-305 (AQ0-39)_. methods used to prepare It is previously app and 2 years old) prior to applying a topcoat of new P does not constitute adequate surface preparation. 4 i M-85 Comanche Peak SSER 9 4
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.,. - - , , , - - . , . - - . , - - nnn- - - -
The TRT reviewed relevant procedures, correspondence from TUEC, and the manufacturer's recommendations. The TRT found that the solvent wipe method is in accordance with the manufacturer's instructions (Imperial Report 462-1-81), provided in TUEC correspondence TXX-4201. The TRT found no evidence that solvent wiping as a surface prepara-tion method under these circumstances was inadequate, improper, or would have any adverse affect on performance of the coatings. This Removal of Debris from Power-Tool-Cleaned Surfaces (AQ0-040). allegation is that certain cleaning requirements are not satisfied. The allegation references TUGC0 procedure QI-QP-11.4-5, Revision 27, paragraph 3.2.2.d, which requires that inspectors verify that blasted or power-tool-cleaned surfaces have been adequately brushed or vacuumed to remove cleaning debris prior to coating application. The allegation states that power-tool-cleaned surfaces were never cleaned, as specified, and that instead these surfaces were blown down with compressed air or wiped with rags. The concern of this allegation is that surfaces might become contaminated with oil, water, or lint by the practice of removing debris with compressed air or rags. In letter TXX-4201, TUEC observes that the same procedure requires inspection to verify that cleanliness criteria are satisfied after removal of cleaning debris. The TRT determined that such visual inspection is generally adequate to detect substantial quantities of oil, water, and lint; however, the TRT considers that inspection is not an adequate substitute for requirements that minimize or prevent such problems from occurring in the first place. The TRT reviewed the requirements of B&R Procedure CCP-30, Revisions 11 and 12, which provide that where the entire thickness of the in-organic zinc primer has not been removed, solvent cleaning shall be employed to remove grease and oil on the surface, typically by wiping the area with solvent-dampened rags. The TRT concurs with this practice. The TRT interviewed ten inspectors and learned that " blowing down" with compressed air had been used to remove cleaning debris, and that solvent wiping was used in a limited manner. The TRT also learned that inspectors will reject any cleaned surface showing contamination by oil, water, or lint. The TRT considers that the effectiveness of any of these cleaning methods, i.e., brushing, vacuuming, blow down, or wiping, will be determined primarily by the skill and diligence of the craftsman. The TRT considers that brushing, vacuuming and blow down are accept-able for removing debri's from surfaces cleaned to bare metal. Air-driven power tools may release small quantities of lubricating oil from exhaust ports, and subsecuent solvent cleaning is an acceptable method to remove lubricating oil. However, solvent cleaning of bare metal with rags should be avoided, so that lint does not adhere to the surface. The TRT concerns about contaminated air supplies are discussed in the assessment of allegation AQ0-17, below. The TRT found that existing l M-86 Comanche Peak SSER 9 l l
procedures do not provide adequate controls to assure that compressed air used to blow down cleaned surfaces is suitably free of entrained oil and water. The TRT was not able to clearly determine the extent to which the practices described by this allegation may have affected coatings. The TRT notes, however, that small quantities of oil, water, or lint undetected on surfaces after final cleaning will usually be revealed during coating application and inspection. The TRT has commented further on this topic in the discussion of allega-tion AQ0-17. Foreign Cleaning Solution (AQ0-41). This allegation concerns the use of a foreign cleaning agent to wipe coated surfaces immediately prior to repairs. It is alleged that the solution used was a hospital disinfectant containing 2% chlorides, a material not allowed by pro-cedures. The implied concern is that chloride-containing materials within the Containment Building might come in contact with stainless steel and result in stress corrosion cracking in the stainless steel. TUEC indicated in TXX-4201, June 22, 1984, that the material in question was Econolemon Disinfectant Cleaner-Hospital Type, manufac-tured by Garland Supply' Company, Fort Worth, Texas. NRC Regulatory Guide 1.54 prohibits the use of any chloride-containing material for cleaning stainless steel, and requires testing of any material which is to be used for this purpose. However, the TRT found no direct evidence that this material had ever been used on stainless steel at CPSES. , The TRT reviewed procedures CCP-30, Revisions 10.and'll, and CCP-30A, Revisions 2 and 3, and found that certain solvents are specified for use prior to repair of primer and topcoat. The TRT interviewed several QA/QC personnel on the use of this unapproved foreign clean-ing solution. The persons interviewed had no knowledge of this mate-rial being used prior to repairs on the steel liner plate, nor of. this solution being used on any stainless steel inside the Contain-ment Building. One individual informed the TRT that this material had been used for a washdown of applied finish coat on the steel liner plate in the Unit 1 Containment Building, and that NCR C-83-01694 had been written as a result. The. interviewed individuals also in-formed the TRT that these areas were always rinsed thoroughly after such cleaning. The TRT reviewed NCR C-83-01694, and found.that it concerned the use of this material over the topcoat only, and there was no indication that the material had been used for repair work as indicated in the allegation. The TRT reviewed the contents of the disinfectant, and determined that it was a detergent which would have no adverse effect upon the finish coat if the coating was thoroughly rinsed after its use. The TRT found no fault with the governing procedures or with the disposi-tion of NCR C-83-01694. The TRT also reviewed B&R procedure CP-CPM 9.2, Revision 0, June 25, 1984, which was provided by TUEC in TXX-4249, dated August 10, 1984. The TRT determined that this new procedure provides the necessary Comanche Peak SSER 9 M-87
controls to assure that foreign chemical materials do not enter the Containment Building area in the future. Cure of Inorganic Zinc Primer (AQ0-43). Allegation AQ0-43 is con-cerned with improper curing of inorganic zinc primers prior to top-coating and with not following procedures to determine if the primer was properly cured. The TRT reviewed the procedures which provided work and inspection requirements for curing inorganic zinc primers. All revisions of B&R procedures CCP-30 and CCP-30A describe methods to perform and inspect the curing of inorganic zinc primers. The current inspection re-quirements are given by TUGC0 procedures QI-QP-11.4-5, Revision 29, and QI-QP-11.4-26, Revision 6, which require cure verification. The TRT found that the methods and requirements provided in the B&R procedures correspond to the recommendations of the coating manufac-turers with one exception, which is discussed under allegation AQ0-44. A generally accepted practice is to accelerate curing by wetting the primer with water after initial drying. This practice was incorpo-rated into the B&R procedures after April 1981, and the TRT found this procedure was generally employed thereafter. The TRT also in-terviewed 11 inspectors who are presently employed at CPSES concern-ing cure verification procedures, and found no evidence to show that curing requirements had not been implemented consistently. In addition, the TRT determined that the backfit program's adhesion testing (see Coatings Category 1) provides an acceptable measure of confidence that inorganic zinc primers applied prior to November 1981, were adequately cured. Performance of the " Nickel Test" (AQ0-44). Allegation AQ:-44 is con-cerned with the " nickel test," used to verify the cure of inorganic zinc primers prior to topcoating, not being properly performed because of oral instructions from QC supervisors to perform the test by lightly rubbing the coating with the coin and to apply just enough pressure to hold the coin in contact with the surface. The implied significance of the allegation is that improper instructions resulted in the topcoating of inadequately cured primers, and that those primers might fail in service. The " nickel test" is a generally accepted technique to verify the cure of inorganic zinc coatings, and has been used since the 1940s. Inorganic zinc coatings harden during curing; the " nickel-test" pro-vides a uniform method of assessing the hardness of the inorganic coating. Experts in the field differ slightly in the methods they recommend for performing this inspection; a consensus standard is currently under development by ASTM Committee 0 01.48. There is a general recognition in the coatings industry that the results of this test can be interpreted very subjectively. TUEC letter TXX-4201 states, in part, that "It is our opinion that the " coin test" method described in CCP-30 and CCP-30A confctms with Comanche Peak SSER 9 M-88
[the manufacturer's] preferred method for the coin test as described in the attached [ Carboline's] letter." The TRT reviewed TUEC's governing procedures, and found that the use of the " coin test" or " nickel test" to verify the cure of inorganic
- zinc coatings was initiated in B&R procedure CCP-30, Revision 7, dated May 7, 1981. Paragraph 4.4.1.1.6 of this revision states, in part, that "[The coating] is sufficiently cured for topcoat when the coat-ing may be burnished rather than removed when rubbed with the flat portion of a smooth edged coin such as a nickel." The statement is maintained substantially unchanged through subsequent revisions, and was incorporated into B&R procedure CCP-30A, Revision 1, and subse-quent revisions.
The TRT found that certain inspector qualification examinations re-quire QC inspectors to describe the method of performing the nickel test. The correct response as given by the prepared answer key, corresponds to the procedures. The TRT found that QC inspectors are not issued controlled copies of CCP-30 and CCP-30A. The nearest available controlled copy of these procedures is maintained in the paint superintendent's office. The TRT requested that some inspectors demonstrate the " nickel" test or to describe it in detail. The TRT learned that inspection person-i nel conduct this test by lightly rubbing the inorganic zine with the i flat of a nickel and then visually examine the coating for the presence of burnishing or any removal. of the coating. The TRT also found that the procedure employed at CPSES differs from the statement given by the manufacturer referenced in TUEC letter TXX-4201. The letter states: "Put heavy pressure on the coin with the finger tips and rub the coin back and forth (8-10 times)." The procedure used at CPSES does not specify the use of heavy pressure or the number of times the surface is to be rubbed. The TRT was not able to establish the extent to which improperly performed " nickel tests" might have resulted in the coating of partially cured inorganic zinc primers. However, the water curing procedure discussed under allegation AQ0-043 is sufficiently effective to eliminate significant concerns of inadequate inspection practices.
- c. Allegations AQ0-18, AQ0-19, AQ0-20, and AQO-51 concern inadequacies in the procedures and methods which were employed in the backfit test program (BTP). A detailed analysis and evaluation of the significance of these allegations in regards to effects upon the BTP is provided in Coatings Category 1. These allegations are considered separately here to provide an evaluation of the adequacy of the governing proce-dures. In evaluating these allegations, the TRT reviewed pertinent revisions of the backfit inspection procedures, TUGC0 instructions QI-QP-11.4-23 (Steel Substrates) and QI-QP-11.4-24 (Concrete Sub-strates), as well as other related inspection procedures.
Visual Defects During Backfit Inspections (AQO-18). Allegation AQ0-18 is concerned with inspection for visual defects such as cracking or i Comanche Peak SSER 9 M-89 i I
blistering; it is alleged that inspectors were not allowed to identify such defects during backfit inspections. In reviewing the backfit inspection procedures and related NCRs, the TRT found that, although some visoal inspection parameters were in-cluded in certain revisions of the procedures, that visual inspection per se was not an element of the backfit inspection program. There-fore, inspectors were not supposed to perform visual inspection as part of the BTP. TUEC's letter TXX-4201 provides the explanation that visual inspections were performed separately in accordance with other procedures that governed ongoing inspection work (i.e., TUGC0 instructions QI-QP-11.4-5 for steel substrates, and QI-QP-11.4-10 for concrete substrates). The TRT has evaluated the adequacy of the visual inspection requirements of those other procedures under sec-tion 4e below. (See Coatings Category 1 for additional information on AQO-18.) Given the fact that the backfit program was instituted as a remedy for previous deficiencies in performance and documentation of inspec-tions, and that these deficiencies included lack of adequate records of visual inspections, the TRT considers that it would have been appropriate to include visual inspection as part of the backfit pro-gram. The lack of such visual inspections created an unnecessary ( complexity in inspection requirements and records which could have been avoided by including all inspection criteria (DFT, adhesion, and visual evaluation for defects) in a single procedure, and the results in a single report for each area. (See Coatings Category 1, AQ0-18. ) Backfit Instructions Vague (AQ0-19). Allegation AQ0-19 concerns TUGC0 instructions QI-QP-11.4-23 and QI-QP-11.4-24, which govern the performance of backfit inspections. It is alleged that these proce-dures are very vague regarding the way the backfit inspections are to be conducted. The TRT reviewed all revisions of these procedures, including QI-QP-11.4-23, Revision 0, November 19, 1981, through Revision 13, April 18, 1984, and QI-QP-11.4-24, Revision 0, February 5, 1982, through Revision 7, April 18, 1984. The TRT also interviewed in-spectors who had performed backfit inspections and examined records of their training in the performance of these inspections. The TRT found that, although these procedures included detailed instructions in certain areas, there were other areas in which the instructions were incomplete or missing. Examples are: (a) These instructions are entitled " Reinspection of Seal Coated and Finish Coated Steel Substrates for which Documentation is Missing or Discrepant" (QI-QP-11.4-23), and " Reinspection of Protective Coatings on Concrete Substrates for which Documentation is Miss-ing or Discrepant" (QI-QP-11.4-24). However, neither instruction provides adequate details regarding what plant areas or items the procedures apply to (i.e., which areas or items have missing or discrepant documentation), or provide details on how such areas are to be identified or by whom. Comanche Peak SSER 9 M-90
.- ~. .- . - - . - - - _ . _ . - ~ - - . _
i instructions These instructions do not provide detailed operat tester. ngmployed for (b) on the use of the two principal instruments ethe Tooke gau inspections: iders that these f of the inspec-1 Despite these procedural deficiencies, Training and quali- the TRT tions by individuals who have been trainefit inspectors, have instruments and implementation of the procedures. tion is report fication ofseparately, inspection personnel, eva ua including back l been reviewed and the TRT's d find no evidence i Coatings Category 7. I ing and qualification of personnel, the TRT coulperformance h o that these deficiencies resulted in improperDuring dequate under- interviews inspections. TRT confirmed that all individuals questioned had an ad the pe standing of the proper use of the instruments an the inspection procedures. J Allegation AQO-20 Adhesion Tests Not Properly Performed (AQO-20 It is alleged . to perform adh concerns the method used by backfit ioninspectors tester. tests of coatings using the Elcometer to adhesto performingcut the the test,coatings aroun l l that the QC inspectors were instructed i not rovided by the manufac-i i the adhesion test dollies (" scribing") pr orand tha l turer of the instrument.these manuf cedures,acturer'sthe instruc l Based upon a review of TUEC's backfit inspection h t this allegation waspro's instruc specification, the adhesion tester manufacture substantially correct: i the tests. dollies, although the manufacturer t d by TUECspecifically to The TRT reviewed the results of onsite lifiedtesting individ-conduc et af demonstrate that scribing or tsnot scribing does for coatings. uals in the industry, including persons curren ing an ASTM standard on the subject iable adverse of adhesion impact testhat fa Based upon this information, the TRT determined around the dollies would not have had any apprect technically im upon the validity of test results and was hadno provided oral The TRT noted, however, that in thisinstructions). to the written case TUEC The directions to the QC inspectors that i were contrary actice to in instructions TRT also noted(2323-AS-31 that TUEC has since and modified the manufacturer's s te pri h the m tute scoring around dollies in compliance w t , instructions. Allegation AQ0-51 concerns t ofthe Phenoline 305 practice of thinning 50/50 Mix Phenoline (AQO-51). 305 (P-305) by a l thinner per gallon. i M-91 4 Comanche Peak SSER 9
material, when dried, to become as brittle as glass and to lose its impact and abrasion resistance. It is alleged that the material became so brittle that it was not possible to obtain a Tooke gauge reading. The TRT reviewed TUEC's application procedures, B&R CCP-30 and CCP-30A, and TUEC's correspondence on the subject (with attached letter pro-viding the manufacturer's recommendations) and determined that thin-ning P-305 up to two quarts per gallon for certain application condi-tions (i.e. , " tie" or " seal" coats, pre-treating sharp edges, and application at relatively low ambient temperatures), was procedurally addressed, in accordance with the manufacturer's instructions, and technically correct. There is no evidence to indicate thet thinning the material in this manner will cause the applied coating to become brittle and to lose its impact and abrasion resistance. The TRT concurred with TUEC's statement in TXX-4201 that any apparent embrittlement which made it impossible to obtain a Tooke gauge read-ing was more likely the result of a dull tip on the gauge than of deficient characteristic of the coating film. This explanation how-ever raised an additional question in regards to backfit inspection procedures and the use of a Tooke gauge with a dull tip. (Tooke ' gauge tips must be sharp to provide a smooth cut and an accurate reading.) Although the TRT did not find any records-to demonstrate that backfit inspectors had received specific instructions in this area, during interviews conducted by the TRT all- backfit inspectors questioned were aware of the consequences of dull tips on Tooke gauges. The TRT also noted that backfit inspection procedure QI-QP-11.4-23 requires a daily confidence check of Tooke gauges, including an examination for evidence of tip wear. The TRT did not find any evidence that the "50/50" mix resulted in any deficiency in the coatings applied or that there was any related improper use of Tooke gauge with dull tips.
- d. Allegations AQ0-24, AQO-42, AQ0-48, and AQ0-49 concern application of coatings to surfaces where they should not have been applied. In each case the concern is that such misapplied coatings could fail, thereby creating debris which could interfere with the proper opera-tion of engineered safeguard systems.
Allegation AQ0-24 concerns Q coatings that have been placed over rusty, scaly, unprepared metal surfaces inside pipe supports made of tube steel without end-caps. The TRT reviewed correspondence from TUEC on this subject, OCA 16,106, and CEL entry 32. The TRT found that entry 32 in the CEL documents 6,000 square feet of CZ-11/P-305 misapplied to tube steel support interiors without proper surface preparation or inspection. The TRT also noted that DCA 16,106, Revision 1, stated that, " Coatings extendir.g into open tube steel members resulting from spray operations performed on the ends and exterior of the member is acceptable." Comanche Peak SSER 9 M-92 I__- _ -_-_ -____-_
f Allegation AQ0-42 concerns coatings which were applied over duct tape and foam rubber in Richmond Inserts. The TRT reviewed correspondence from TUEC on the subject, DCA 12,374, and the CEL. The TRT found that DCA 12,374 downgrades all coatings on Richmond Inserts to speci-fication 2323-AS-30 (Non-Q), and that entry 30 in the CEL documents 2,258 square feet of Richmond Inserts in Reactor Building 1. Allegation AQO-48 cnncerns coatings which were applied over " seismic joints" which were filled with foam and were not to be coated. The TRT reviewed correspondence from TUEC on this subject and the CEL. The TRT found that the joints described in the allegation were in fact expansion joints rather than seismic joints, that the total area involved was approximately 125 square feet, and that this item had i been entered into the CEL. In all three cases, the TRT found that TUEC's estimate of the size of .! the area involved was acceptably conservative, and that entry of the l item into the CEL provided an acceptable resolution of the problem. In evaluating concerns with misapplied coatings, the TRT reviewed the governing specification and all pertinent procedures and instructions. The TRT found that although specification 2323-AS-31 provides a list-ing of items which require coatings and items which do not require
- coatings, nowhere in the specification or procedures are there any instructions or requirements for protecting items which are not to j be coated or for inspection activities to assure the adequacy of protective measures which are taken. The TRT confirmed this finding through interviews with coatings quality engineering personnel, each of whom stated that no procedural requirements for masking or inspec-tion of masking exist, and that any misapplied coatings which are detected during other inspections are handled on a case-by-case basis.
t Allegation AQ0-49 concerns overspray of coating materials onto coated surfaces that had previously been inspected. It is alleged that this has been allowed and is commonplace. The TRT reviewed correspondence i on this subject from TUEC, reviewed pertinent procedures, and examined i completed coating work in Reactor Buildings 1 and 2. The TRT concurs, l in principle with the evaluation of this issue provided by TUEC in 1 letter TXX-4201, i.e., overspray is a common phenomenon during the spray application of coatings and is not harmful, provided that ade-quate methods are employed to detect and correct any excessive or detrimental overspray which occurs. The TRT found no evidence of excessive uncorrected overspray in its examination of finished coat-ing work. However, the TRT noted that the governing procedures do not provide guidelines or instructions for the protection of finished work. i ! The TRT found that the lack of instructions or methods for masking l and protecting items not to be coated, and the lack of inspections for adequate masking, were the cause of the specific deficiencies described in these allegations. The TRT also found that these proce-dural deficiencies would allow continued and possibly undetected application of coatings to non-specified surfaces which could result Comanche Peak SSER 9 M-93 4
in failure of such misapplied coatings as well as potential damage to sensitive plant equipment fron coating activities.
- e. Allegations AQ0-17, AQ0-23, AQ0-31, AQ0-45, and AQ0-58 are all concerned with the performance of quality control inspections for coating work at Comanche Peak. These allegations concern inspections which were not performed or were performed improperly, and which were therefore inadequate to correctly assess the acceptability of the coating work and the applied coatings.
Air Acceptance Invalid (AQ0-17). Allegation AQ0-17 is that the in-spections of compressed air cleanliness prior to conventional spray application of coatings are invalid because production personnel- in-sert cigarette filters into the air line immediately ahead of the test point without the knowledge of QC inspectors, and then remove the cigarette filter after the test. It is further alleged that construction and QC management were aware of the practice. The sig-nificance of this allegation is that the use of compressed air con-taining entrained oil and moisture for sand blasting, blowdown, and conventional spray application, will degrade the applied coatings. The TRT interviewed inspectors, QC management, and reviewed TUEC's response to this concern given by letter TXX-4201. The TRT found that the alleged practice did occur, but that it has been stopped. The TRT also found that air supply equipment for Unit 1 was' replaced in September 1983, and that the same equipment was moved to Unit 2 in August 1984. During its review, the TRT tested the air supply at a Unit 2 coating operation and found it' acceptable. The TRT found that QC inspectors who perform this check periodically reject compressed air cleanliness, and that replacement of filtering elements is then performed and is adequate to correct the adverse conditions. In the experience of the coatings industry, oil and water in compressed air will not have significant adverse effects on coating performance unless present in sufficient quantities to con-dense. This observation has resulted in the generally accepted test method, used at CPSES, of holding a clean white blotter in the com-pressed air stream and then examining the blotter for traces of oil or moisture. CPSES procedures require a-30-second blotter test, which is at the lower limit of.the range of periods for which this test is typically conducted. Oil and water present in sufficient quantities to condense upon the applied coatings are ofteri readily visible to a trained observer without benefit of the blotter test. Unacceptable quantities of water will shorten the pot life of coatings in pressure pots, especially for inorganic zines, or cause certain discolorations. Water deposited on surfaces by blasting or blowdown will cause rapid " rusting" dis-coloration. Unacceptable quantities of oil deposited on surfaces by blasting, blowdown, or spray atomization will cause a visible defect known as " fisheyes," or a sliding of the coating known as " creeping," and may produce discolorations of the coating film. Minute quantities Comanche Peak SSER 9 M-94
of insoluble oils may be retained within coating films without ap-Minu preciable adverse effect. out of the coating film. The TRT examined the current revisions of theThe applicable TUGC0 in-TRT considers spection procedures from the QI-QP-11.4 series. that the procedures provide generally acceptable methods for a pro-perly trained inspector to identify and verify correction of such defects in the applied coating which might result However, the TRTfrom thethat found useproce-of inadequately clean compressed air. dures do not provide rigorous methods to limit the recurrence of the deficiencies. Although this allegation Coatings QC Program Inferior (AQO-23).it is concerned specifically with represents a subjective judgment, It is alleged that the
" standard" inspection methods and techniques.
Comanche PeakTheCoatings QC inspection program is in example given in the allegation is the are not used at CPSES. American Society for Testing and Materials (ASTM) tape adhesion test which, it is alleged, was used regularly at another site, but not allowed at CPSES by one of the QC lead inspectors. The TRT reviewed TUEC's governing procedures, specifications, cor-respondence from TUEC on the subject, relevant standards, and commonThe practices at other nuclear power plant projects. specific coatings inspection practices and procedure terpretation of regulatory and industry standard requirements, and on how these requirements were translated into a site-specific quality control program. In terms of the types of inspection techniques, instruments, and methods employed, the TRT found that CPSES was not appreciably different from other sites as to inspections addressed by the coatings QC program. Specifically in regards to not allowing the ASTM tape adhesion test, the TRT does not consider that this demonstrates that TUEC's coatings The TRT did not find that the ASTM tape test QC program is inferior. was regularly or extensively used at other sites, n test. When adhesion testing is performed at nuclear sites, the commonly used method is employing the Elcometer adhesion tester, which is the method addressed by ANSI NS12 for t by CPSES specification 2343-AS-31 and the gove inspectors not to use an unspecified alternate method. This allegation concerns the inter-Limited Access Areas (AQO-31). l pretation of CPSES OCA Revision 2,13,140, February 21,which 1983, of addre access or which were inaccessible.this DCA, which has been sion 2) of specification 2323-AS-31, specifies that limited-access M-95 Comanche Peak SSER 9
4 4 4 areas shall be prepared to Steel Structures Painting Council (SSPC) specification SP-10 (near-white metal blast) or equal, if possible, with a minimum requirement of SSPC-SP-6 (commercial blast) or equal. The DCA further specifies that inaccessible areas shall be treated on a "best effort" basis and that QC inspection on inaccessible areas is not required. The DCA includes drawings which illustrate and define
" limited access" and " inaccessible" areas. It is alleged that QC management interpreted the SSPC-SP-6 minimum requirement for limited access areas to mean "do the best you can" and stated to QC inspec-tors that if they could not get to an area, not to worry about it.
The TRT reviewed all revisions of DCA 13,140, specification 2323-AS-31, and correspondence from TUEC on this subject. The TRT recognizes
- that limited access and inaccessible areas exist in all nuclear power plants, where it is not possible to perform coating work operations and inspection activities in accordance with normal specification requirements. The TRT considers that the requirements in DCA 13,140 provide appropriate instructions for performing and inspecting sur-face preparation work in such areas. The TRT noted that the changes from Revision 0, to Revision 1, to Revision 2 of this DCA provided progressively more explicit criteria for determining the classifica- +
tion of an area or item as a limited-access or inaccessible area. The TRT could not confirm that the alleged statements were made by QC management. The TRT considers that any areas which have been classified as in-accessible or limited-access areas in accordance with this DCA, do not meet the normal specified requirements for Service Level I coat- ! ings, and should, therefore, be entered into the CPSES CEL. However DCA 13,140 makes no provision for entry of these areas into the CEL . as a justification for the downgraded requirements. In TXX-4262, August 21, 1984, TUEC provided an estimate of approximately 6,100 square feet of surface area classified as inaccessible or limited access, and the TRT found that this figure was acceptably conserva-tive. Further discussion of the CEL is provided'in Coatings Cate-gory 6. Reinspection of Repairs (AQO-45). This allegation concerns the re-inspection of repairs that are performed to remedy defects in applied coatings. It is alleged that some coating repairs at CPSES were , never reinspected and that other coating repairs were not given the ' same type of final inspection that would have been performed for , regular production work. The implied significance of this concern '; is that repairs which are not reinspected properly may have defects or may sustain damage which would not be detected and corrected. The TRT reviewed the current governing procedures identified by TUEC, , TUGC0 QI-QP-11.4.5 and QI-QP-11.4-26, and correspondence on this ; subject from TUEC to the TRT. The TRT also reviewed all revisions of TUEC's procedures which have governed coatings repair work in the past, including QI-QP-11.4-5, QI-QP-11.4-10, QI-QP-11.4-26, and , QI-QP-11.4-27. l f Comanche Peak SSER 9 M-96
. . - _ . - - .- - -. ~- . _ - . ~
l i l The TRT found that the requirements for reinspection upon completion of repair work were not Anclearly example ofaddressed an adequate in all ca were, generally, adequately addressed. procedure address These sections provide revision 6, paragraphs 2.10.2 and 2.10.3. specific inspection requirements for repair of both major in and minor defects, and include requirements to " perform inspection This sec[ tion].. 2.9" (" Finish Coat Final Acceptance Inspection"). procedure also includes provisions An example of an for a " Steel 1 ensure completion of all required inspections. l inadequate procedure addressing inspection of repairs is found in 2 QI-QP-11.4-5, Revision 29, May 4, 1984, in which paragraphs 3 j 4 3.7.3 do not provide equivalent Similar requirements defici- and d l with paragraph 3.6 of that procedure) for repairs. The i encies occur in earlier revisions of the governing procedures. l TRT found that both in current and historical inspection procedures, l' clear instructions were not always provided for reinspection of i repairs. The TRT noted that TUEC has made a distinction between the insp f requirements for " minor" versus " major" repairs, and this issue wasR i discussed by TUEC in letter TXX-4201, June 22,1984. i specific issue, the TRT found that the distinction drawn by TUEC w l appropriate, noting that it is consistent with standard industry prac- < j tice to perform final inspections of minor repairs (such as "touching i up" pinholes) while repairs are being made. l l i The TRT also noted that TUEC has stated in correspondence and inte i ! views that a final visual inspection performed b
. Revision 1, provides assurance that all coated areas, includingThe TR f repairs, receive a final visual inspection.
l procedure and found it inadequate for the purpose stated abov
; that it does not include visual inspection methods, criteria, or j adequate documentation of same, and does n activities.
However, during its visual examinations of coated surfaces within l' CPSES Unit 1 Containment Building, the TRT a damage to repaired areas. L This allegation concerns
! Inadequate Lighting for Inspection (AQ0-58). in-i procedural requirements for illumination used to perform It coa ,
l spections and cites TUGC0 instruction QI-QP-11.4-1 as an el ' l is alleged that the instructions provide for inadequate lighting. I l i The implied concern is that inspections perfor l and may not identify defects which are present. 1 i M-97 Comanche Peak SSER 9 l
l The TRT reviewed the governing procedures and specifically examined QI-QP-11.4-1, Revision 20, March 5,1984, which is currently in effect. Paragraph 3.0 of that instruction states: " Visual inspec-tion of surfaces as addressed by this instruction shall be made at approximately an arms length from the surface being inspected. The area of inspection shall be adequately lighted during the inspection activity. Adequate lighting is defined as the minimum light produced by a two (2) D-cell battery flashlight. Flashlight shall be held perpendicular to the surface during visual inspection." The TRT considers the language of this paragraph to be vague, technically inappropriate, and contrary to recognized standard practice for per-forming visual inspections of coatings. Proper practice involves the examination of coatings work from dif-ferent angles, and sidelighting the surface to identify certain defects (e.g., hackles or protrusions in blasted steel, and blisters, craters, and runs in applied coatings). It is, therefore, important that visual inspections require the use of a movable light source and side lighting of the surface as needed (rather than a requirement to maintain the light " perpendicular" to the surface). The inspector should also have the freedom to vary his distance from the surface (both for overall examinations for shadowing and color uniformity and for closer examination of questionable areas), rather than being . restricted to an " arms length from the surface." The language in the i paragraph quoted above regarding light intensity can easily be mis-interpreted to imply very weak light or no light, but even if it means the light provided by a fully charged, fully functioning two-D-cell flashlight, this may not be adequate in all cases. The TRT finds that the allegation was substantiated in that TUEC provided inappropriate and inadequate requirements for pe-forming visual inspections in paragraph 3.0 of QI-QP-11.4-1,' Revision 20. Rigorous adherence to the limitations of this paragraph would severely restrict an inspector's ability to adequately perform visual inspec-tions and to identify coatings defects.
- f. Allegations AQ0-34, AQ0-36, and AQ0-62 all concern coatings trace-ability in that they address alleged deficiencies in TUEC's handling, storage, and use of coating materials that would render traceability of applied coating materials to DBA tested batches indeterminate.
These three allegations are evaluated and discussed in detail in Coatings Category 3. They have been included for discussion in this report, as well, for consideration of their procedural aspects. A brief summary of these allegations and the TRT's evaluation of each follows: ANSI Requirements Not Met (AQ0-34). This allegation, concerning non-compliance with ANSI N45.2.2 for coatings material storage, was sub-stantiated but not considered to be significant because compliance with that standard is not required by the CPSES FSAR. Coating Materials Traceability Not Maintained (AQ0-36). This allega-tion, concerning failure to maintain coating materials traceability, Comanche Peak 55ER 9 M-98
was substantiated in regards to areas included in the backfit test program for which previous records were missing or discrepant. Contaminated Paint (AQ0-62). This allegation, concerning contaminated materials being applied, was not substantiated to the extent that the TRT did not find evidence that contaminated materials had been applied. However, in the TRT's generic review of TUEC's system for control of coating materials (as reported Coatings Category 3), the TRT deter-mined that the procedurally addressed system was inadequately rigorous in several respects. Brown & Root (B&R) procedure CP-CPM 8.1, Revision 1, " Receipt, Storage, and Issuance of Items," provides requirements for control of stored coating materials prior to issuance to the paint department. However, once materials have been issued from the site warehouse to the paint department, the only storage control riquirements are those provided by CCP-30, CCP-30A, and CCP-40. These procedures address storage temperature ranges and the location of "Q" materials in the paint storage warehouse that are segregated from non-Q materials, and little else. These procedures do not address any physical inventory control measures or requirements for moving materials in and out of the paint warehouse. They do not describe controls to assure the continued segregation of Q and non-Q materials when they are moved in and out of the warehouse and are transported to and from different locations on site. TUGC0 procedure QI-QP-11.4-17, Revision 6, " Surveillance of Storage and handling of Protective Coatings," provides for monthly storage inspection of storage facilities by QC. However, it imposes no re-quirements for more frequent checks to verify that physical control is maintained in the interim or for control by QC of materials during transport on the job site or for storage at other job site locations.
- 5. Conclusions and Staff Positions: Based upon its review, the TRT has reached the following conclusions:
- a. For allegations concerning DBA qualifications (AQ0-01, AQ0-02, AQ0-03, AQ0-04, AQ0-05, AQ0-06, AQ0-11, AQ0-12, and AQO-15):
The individual items described in the allegations are not significant because they concerned only minor surface areas, were properly cor-rected by TUEC, or were not substantiated. However, Coatings Cate-gory 2 also addresses TRT technical concerns related to DBA qualifica- ' tion of the major coating systems which are beyond the specific scope of these individual allegations. The significance of the findings for these allegations from a proce-dural perspective is that TUEC's failure to properly address special-case coating systems for overlaps and repairs has resulted in some applied coating systems that are neither qualified nor proven to be technically viable. Comanche Peak SSER 9 M-99
- b. For allegations concerning procedures and instructions which provide technically incorrect directions:
AQ0-07. While the TRT found the disposition and repair for the NCR involved in this particular allegation to be generally acceptable, the TRT found that TUEC's procedures governing application of concrete materials as detailed above, were not in compliance with the manufac-turer's instructions. The significance of this finding is that some materials applied at CPSES may not perform adequately in service or under DBA conditions. AQO-08. The TRT has not substantiated this allegation. AQ0-10. The TRT concludes that TUEC's procedures for power tool cleaning surfaces for touch up work, and inspection of this work, were inadequate to assure acceptable roughening of the surface. The significance of this finding is that some materials applied to these areas may not exhibit adequate adhesion. AQ0-38. The TRT concludes that this allegation is not substantiated. AQ0-39. The TRT concludes that this allegation is not substantiated. 1 AQ0-40. The TRT concludes that TUEC's procedures for removal of debris from power-tool-cleaned surfaces were inadequately rigorous, but unlikely to result in undetected defects in applied coatinfr. l Therefore, this allegation is not significant. AQ0-41. The TRT concludes that the use of the foreign cleaning agent as it occurred at CPSES did not have any adverse effect upon applied } coatings, and that current procedures adequately control future use of such materials. Therefore, this allegation is not significant. AQ0-43 and AQ0-44. The TRT concludes that the concerns raised by these allegations did not result in irnproper curing of inorganic zinc at CPSES. Therefore, these allegations are not significant.
- c. For allegations concerning backfit inspections:
AQ0-51. The TRT has not substantiated this allegation. AQ0-18, AQO-19, and AQO-20. As discussed above, backfit inspection procedures were inadequate in certain respects, however the TRT did not find that tMse inadequacies would have any significant effect upon the data ggnerated by the backfit program. (See Coatings Cate-gory 1.) i d. For allegatioht concerning coatings applied to areas where they should not have been applied: l The TRT found that the areas described in AQ0-24, AQ0-42, and AQ0-48 have been entered into the CEL and therefore these allegations are not significant. AQ0-49 was not substantiated. However, the TRT Comanche Peak SSER 9 M-100
found that procedures for protective measures to be taken for items not to be coated were inadequate, resulted in defects which could have been avoided, and created the possibility of unnecessary damage to other plant items. (The TRT did not find evidence that such damage occurred.)
- e. For allegations concerning the performance of QC inspections:
AQ0-17. The 'RT concludes that this allegation is substantiated in regard to the insertion of filters during air acceptance tests, an action which invalidates those test results. The TRT did not find any evidence that this resulted in defects in the work. However, this allegation is significant in that it represents interference with the proper performance of a QC inspection activity and therefore renders recorded results for air acceptance tests, and possibly other inspections, indeterminate. AQ0-23. The TRT concludes that this allegation is rot substantiated. AQO-31. The TRT concludes that this allegation is not substantiated. AQ0-45. The TRT concludes that this allegation is substantiated in regard to inadequate procedural address of reinspection for repair work. The significance of this finding is that there may be unde-tected and uncorrected defects in repaired areas, although the TRT did not observe any such cases. The TRT found that TUEC's provisions for a final engineering walkdown inspection are not an acceptable remedy for this concern. AQ0-58. The TRT concludes that this allegation is substantiated in regard to inappropriate instructions and requirements for illumina-tion during visual inspections. The significance of this finding is that visual inspections may not have been performed properly; consequently, there may be undetected defects in the applied coatings.
- f. For allegations concerning traceability (AQ0-34, AQ0-36, and AQO-62):
A detailed evaluation of the significance of these allegations is provided in Coatings Category 3. The significance of these concerns in regards to procedures is that inadequately rigorous procedural address of coating material control measures may have resulted in unnecessary loss of traceability. The significance of.the substantiated allegations discussed above, in regards to the TRT's evaluation of TUEC's procedures and instructions governing coating work, is that those procedures were inadequate and inappropriate in many respects. This resulted in defective work in certain instances, which led to an unnecessarily large area of exempt coating work being placed in the CEL. These inadequate procedures also resulted in inadequate performance of inspections in certain instances where proper inspections could have prevented or corrected defective work. The TRT concludes that these procedural deficiencies demonstrate that review and approval of procedures and instructions by TUEC was inadequate to detect Comanche Peak SSER 9 M-101
1 f i and correct these deficiencies. Inadequate performance of procedures review and approval activities indicates that the personnel who performed those activities may not have been qualified to make the correct technical and quality evaluations of those procedures. These deficiendies, although not of safety significance in the coatings area must be considered in J evaluating the effectiveness of the overall-QA/QC program. Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original concerns and to obtain any additional comments from them. A summary of the followup interviews is included in Section 2.2.3 of this appendix.
- 6. Actions Required: The TRT found deficiencies in procedures and instruc-tions for coating work and related inspection activities, during the j construction phase, which rendered them inappropriate or inadequate for determining satisfactory accomplishment of important activities. The TRT also found that TVEC's procedure review and approval system was inadequate to detect and correct these deficiencies.
Accordingly, TUEC shall make the necessary changes to the procedure review and approval system to assure review and approval by technically qualified individuals to prevent recurrence of the types of deficiencies discussed above, and to assure procedures are reviewed for consistency and clarity. This revised review and approval system shall be applied for the issuance and revision of all procedures which will govern future coating work, 1 inspection, and testing at CPSES consistent with the guidelines of j Appendix L. . 4 l i i f i Comanche Peak SSER 9 M-102
Allegation Category: Protective Coatings Sa, Inspection Reports 1. Parts of AQ0-14, AQ0-25, AQ0-49, AQ0-50, AQ0-52,
- 2. Allegation Number:
AQO-53, AQO-5/ and AQ0-59 Characterization: It is alleged that: 3.
- Inspection Reports-(irs) can be dispositioned by anyone (AQ0-14b).
Instead of writing NCRs, irs must be written as unsatisfactory (AQ0-14b). Nothing prevents items identified on unsatisfactory irs fr A seal coat was accepted by QC personnel prior to the finish coat being applied when the seal coat should have been rejected (AQ0-25). Coating material is oversprayed into areas that were Coatings have been applied without the benefit of quality control inspection (AQ0-50 and -52).
- QC inspectors were denied the opportunity to write requests for information or clarification (RFIC) (AQ0-53).
- Coatings were applied over filth (AQ0-57).
Substandard coatings were accepted by QC inspectors (AQ0-59). To address the allegations concerning
- 4. Assessment of Safety Significance:
~
disposition of irs for protective coatings, the NRC Technical Review Team The TRT (TRT) conducted an examination of the IR system if in this are which indicates that irs are the primary methoo Theused TRT to documentofsat examination s ac-many tory and unsatisfactory coating inspections. irs shows that coating problems usually occur during onsite applications J and inspection, a time when repairs can easily be identified, documente and made through the existing IR system. TRT found that irs which document problems that cannot be corrected using standard (NCR). repair procedures are then documented by a n e IR system and its implementation to be satisfactory. l In assessing the allegations that com-Improper Use of irs (AQ0-14b). 4 l
~
prise AQO-14, the TRT interviewed Coatings QC personnel Revision 22.and found tha 4 they avoided writing NCRs because of Procedure QI-QP-11.4-5," Nonconformi i Paragraph 3.9 of this revision states: TUGC0 Procedure 7 be reported on an IR in accordance with CP-QP-18.0." lists three methods by wh I 1 CP-QP-18.0, Revision 12, July 19, 1983, are closed: (1) all items are satisfactory; (2) all unsatisfactory items have been repaired and reinspected and found satisfactory; or (3) an NCR I has been issued for unsatisfactory items. M-103 Comanche Peak SSER 9
I i The TRT reviewed QI-QP-11.4-5, Revision 22, which was in effect during the i period of the allegations, and found that Paragraph 3.9 was not understood by some Coatings QC inspectors. However, QI-QP-11.4-5 referenced CP-QP-18.0, j Revision 12, which in the opinion of the TRT PC Group contained adequate
- methods for closing an IR. Revision 29 of QI-QP-11.4-5 clarifies when NCRs L should be prepared. Discussions with Coatings QC inspectors on the current
! revision of QI-QP-11.4-5 (Revision 29) indicate that they are satisfied with
, this revision and understand it. TRT interviews with the QC inspectors ! indicated that even with the old revision of QI-QP-11.4-5, QC did not ignore 4 nonconforming conditions. Review of site procedures by the TRT indicated } that irs cannot be dispositioned "by anyone," as alleged. The only differ-j ence between Revision 12 and Revision 20 (the current revision was issued
- in September 1984) of CP-QP-18.0 is-that Revision 20 specifies that a DCA i may also be issued which makes the unsatisfactory condition acceptable.
1 The TRT reviewed the Texas Utilities Electric Company (TVEC) Paper Flow l l Group (PFG) system which issues protective coating inspection report (PCR) j numbers and tracks all incomplete work packages. This system was initiated January 1, 1984. The PFG collected all unsatisfactory irs prior to January 1984, and developed work packages which documented all coating work, including repairs in a given plant area. The TRT performed a sample < review of approximately 100 unsatisfactory irs issued between December 1, t 1982 and January 1,1984, by which time most coating work except for repairs was completed in Unit 1. The.TRT found that these unsatisfactory irs had been or were being properly processed and tracked. Work packages and irs were updated on a daily basis-by entering deficient, discrepant, ] and completed work packages on a computer _ list. The implementation of the PFG system and protective coatings work packages prevented items identified
- - on an IR from becoming lost, and therefore not corrected. The PFG system i appears to work effectively for the protective coatings area.
t
! Unacceptable Seal Coat (AQ0-25). Allegation AQ0-25 concerned a seal coat
- on the liner plate in Unit 1 outside the Skimmer Pump Room being accepted
} when it should have been rejected. The alleger states that stains on the liner were, in his opinion, " unacceptable per procedure." i TRT interviewed the QC supervisor involved who indicated that the stains i were acceptable by Procedure QI-QP-11.4-26, Revision 6. He stated that i the liner was wiped with solvent and water, that the QC inspector involved appeared to be satisfied with the work, and that no pressure was placed on the QC inspector for him to accept the work'as done. Review of the related !- irs by the TRT confirms that the inspector did, in fact, sign-the IR as satisfactory. The TRT could not verify that the QC inspector was coerced. \ l The TRT determined that stains on organic coatings not removed with water and solvent wiping are acceptable. A TRT observation of the area indicated
< that it had been topcoated and accepted.
1 Overspray (AQO-49). Allegation AQ0-49 involves alleged overspray into ] areas that had previously been inspected, which allegedly was allowed and
- was commonplace.
i The TRT review of coating procedures for concrete and structural steel found'that overspray/ dry spray were adequately addressed in the procedure Comanche Peak SSER 9 M-104 1
h timeframe of the allegations. t s each of whom revisions made before, during, and after t e corrected The TRT then randomly interviewed coatings QA/QC inspec or , in accordance stated that when overspray was encountered, it was with applicable QC procedures. adhesion Excessive overspray of primer,the TRT review if not of QC andcan cause4-2 corrected, problems when a final coat is ap,;1ied; however, moved, while a minor am construction procedures (QI-QP-11.4-5, 4-10, that overspray for primer was supposedItto be re Small amounts of final of overspray was acceptable for the l final is not uncommon for coat. or topcoat overspray pose no adhesion probdems. y applications, and CPSES this issue. overspray to occur on adjacent areas during spracoat Allegations Coatings Applied Without QC Inspection (AQ0-50 and AQ0-52_1d without the bene I AQ0-50 and AQO-52 address protective coatings applie of QC inspections. instances inHowever, which the C inspection. The coatings TRT hadreviewed been applied several without NCRs andQsuch this allegation, appropriate found as a documente The alleger gave only general locations affected specificby location. tact the alleger by tele-hanger located on the steellleger liner, rather than a1RT m to call the TRT phone The foralleger more specific information.and never responded. s a co-worker of the a collect. forwarding address from the post office, but had no succes . QC inspection program The TRT review of NCRs indicated thatnonconformances identified the existing which had was working in that it documented andAllAllareas identified revisions of proce-were reworked or d not received QC inspections. llowing the allegation, positioned in accordance with site procedures. inspection must be dures CCP-30 CCP-30A, tive and coatings.CCP-40, prior to performed during in-process application i of pliedprotec without QC Although there have been incidents of coatings be ng apblished to m inspections, adequate procedural measures were esta their recurrence. Information or QC Inspectors DeniedAllegation Opportunity AQ0-53 of WritingallegesRequests for that QC inspe Clarification tors were not(RFIC) (AQO-53J. permitted to write RFICs. ths in mid-1982) a t cted QC inspectors C indicates that for a short period of time (2 to 3 that they not to write RFICs.TRT interviews with presentCQC inspectors indic inspectors management. l i have write them.not been denied the opportunity to writeR er a question the are writing RFICs.fication of a procedure or specification or to answ inspector may have on an item. are procedurally mandated. M-105 Comanche Peak SSER 9 1
The TRT found that the allegation was substantiated. However, there are no procedural requirements for RFICs; therefore not writing them does not violate any requirement. Coatings Applied Over Filth (AQ0-57). Allegation AQ0-57 alleges that at the 860-ft elevation of Unit 2, in the room directly off the elevator, coating was applied to surfaces which were covered with filth, weld spatter, tobacco juice, and other unsuitable material. The specific area was in Room 163, rod position indication (RPI), elevation 860, Unit 2 Reactor Building, a location documented by NCR-C84-0812 and PC 45291 and attachments. These documents refer to coating being applied to an adjacent area in this room where surface preparation or cleaning had not yet been done. The TRT located the room in question and observed that repair work was in progress on the unacceptable area. The allegation is substantiated; however, proper documentation exists, and proper disposition of the problem was made. Acceptance of Substandard Coatings (AQO-59). Allegation AQ0-59 alleges that.QC inspectors accepted substandard coating on the liner plate below and above the polar crane rail at azimuth 270* to 0 in the Unit 1 Contain-ment Building. The TRT examined all protective coating (PC) inspection reports related to the area from azimuth 270 to 0 between the 905-foot and 940-foot elevations on the containment steel liner plate. The review indicated that many irs noted unsatisfactory conditions, which were later corrected. None of the inspection reports indicated that QC inspectors accepted substandard coatings. Only one QC inspector involved in writing these irs was still working at Comanche Peak. The TRT interviewed the remaining QC inspector, but he gave no indication that QC inspectors had accepted substandard work. The TRT i also interviewed nine current on-the-job QC inspectors at CPSES, all of whom stated that QC inspectors do not accept substandard work.
' 5. Conclusion and Staff Positions: Based on the TRT review of the TUEC inspection report system and their implementation of it,'as well as on the TRT investigation of alleged improper use of irs, unacceptable seal coat, overspray, coatings applied without QC inspections, coatings applied over filth, and acceptance of substandard coatings, the TRT concludes that reviews of all issues involved indicated in each case that proper documents existed and corrective action was initiated. Allegations AQ0-50, -52, -53, and -57 were substantiated, but the proper documentation and-appropriate corrective actions were taken. .The TRT concludes that these allegations
~
have neither safety significance nor generic implications. Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original A summary of concerns and to obtain any additional comments from them. the followup interviews is included in Section 2.2.3 of this Appendix.
- 6. Actions Required: None. ;
l Comanche Peak SSER 9 M-106 i i
--_---_x___--____________-____
- 1. Allecation Category: Protective Coatings 5b, Nonconformance Reports
- 2. Allegation No.: AQO-7 and AQO-14
- 3. Characterization: It is alleged that: (1) the disposition of noncon-formance report (NCR) C83-01986, which dispositions cracking and flaking of concrete coating *, will not remedy the problem (AQ0-7); (2) after an NCR is written, anyone can sign off on it (AQ0-14a); (3) NCRs cannot be written and IRo must be written as " unsatisfactory" (AQG-14b); and, (4) a past QC supervisor voided many NCRs (AQ0-14c).
- 4. Assessment of Safety Significance: To assess the allegations concerning Hisposition and approval of NC E for protective coatings, the NRC Technical Review Team (TRT) conducted a. generic examination of Texas Utilities Electric Company's (TUEC's) NCR' system for coatings.
The procedural definition of a nonconformance is a deficiency in charac- , teristic, documentation, or procedure which renders the quality of an item unacceptable or indeterminate. The TRT reviewed Texas Utilities Generating Company (TUGCO) Procedure CP-QP-16.0, "Nonconformances." This procedure describes the system for identifying, resolving, and closing out noncon-formances. The TRT specifically reviewed revisions 8 and 14 of procedure CP-QP-16.0, because they were in effect during the period bracketing the
! - allegations.
ToauditTUEC'simplementationoftheNCRsysteminthekoatingsarea,the TRT reviewed approximately 30 completed NCRs to determine if their disposi-tion was adequate. The TRT noted that several of these NCRs showbd insuf-ficient documented engineering justification. Since these NCRs mainly involved design basis accident (DBA) qualification of coatings or coating traceability, they were assessed in Coatings Categories 2 and 3, which discuss the generic problem that TUEC's NCR system does not specifically require, for significant condicions adverse to quality, identification of the cause of the problem or the necessary corrective action taken to pravent the problem's recurrence. The TRT also found that NCRs for protective coatings were not trended pro-
. pyly. -The present Quality Trend Analysis Reports, which were issued in accordance with TUGC0 procedure CP-QP-17.0, " Corrective Action," combine coating NCRs with other civil engineering NCRs. Combining the,e NCRs in the trending analysis provided a composite trend for protective coatings which was neither representative nor accurate. Therefore, the trsnding system was inadequate to identify and contiol recurring deviations or.
overall increases in deviatio'ns for protective coatings. NCR C83-01986 pertains to the cracking and flaking of concrete coatings systems (Nutec-11, -11S, -1201). Allegation AQ0-7 is coricerned with the disposition section of this NCR, which states that " cracking of coat'ings is due to excessive stresses in the coating during drying and curing" and. accepts reworking of the affected areas as the proper disposition. The alleger believes that the disposition of this NCR was inadequate, and that repairing these cracks will not remedy the condition which caused the cracks. (This subject is discussed in detail'in Coatings Categories 2 and 4 and under AQ0-06 and AQO-U7.) , , Comancha Peak SSER 9 M-107 r e - -* '*<y-y- w- * - *
- r+
The TRT reviewed NCR-C83-01986 and the manufacturer's (Imperial) applica-tion bulletin. Both documents indicate that the disposition of the NCR was adequate. During the cure process, chemically converted organic coatings develup internal stresses, due to shrinkage, which may cause cracking and peeling. Proper application will prevent a recurrence of this condition. The TRT reviewed TUGC0 procedures QI-QP-11.4-27 and CCP-40 and verified that the procedural requirements assured proper application and emphasized
~
control of film thickness. The TRT also performed a walkdown inspection of the Unit 1 Containment Building. The TRT found no visual signs of stress cracking and observed that the affected areas apparently had been repaired. The TRT reviewed TUEC's methods for incorporating vendor coating recom-mendations and information from vendor application bulletins into their coating procedures. (These findings are reported in Coatings Category 4.) To assess allegation AQ0-14a, the TRT reviewed TUGC0 Procedure CP-QP-16.0. This procedure indicates that NCRs are the primary means of documenting nonconforming conditions that cannot be corrected by standard procedural repair or nonconforming conditions which are indeterminate. The TRT reviewed revisions 8 and 14 to CP-QP-16.0, which bracketed the time period of the allegation. These revisions required NCRs to be prepared, reviewed, dispositioned, and signed off by personnel authorized to verify closure of an NCR. Procedure CP-QP-16.0 states that "QC discipline supervisors shall ensure that the NCR disposition work items are witnessed by QC inspectors." The procedure further states that "QA/QC supervisors shall sign the verifica-tion block." TUGC0 procedure CP-QP-16.0, "Nonconformances" (Paragraph 3.2.7) also requires necessary approvals for those portions of NCRs that are affected by revisions. However, in reviewig; actual NRCs, the TRT found that only the original handwritten copy of an NCR is signed and dated by the QC inspector. On subsequent, typed copies of NCRs, the name of the reporting QC inspector is typed on the NCR and is neither initialed nor dated. The alleger may have been referring to these typed NCRs. On September 10, 1984, the TRT telephoned the alleger to obtain further information on allegation AQ0-14c. The TRT asked the alleger what was meant by the term " voided," i.e., that an NCR was cancelled but remained in the record file, or that it was destroyed and there was no longer a record. The alleger was not certain whether any NCRs were physically destroyed, but believed that the disposition of one NCR was not adequate. Therefore, the alleger considered that NCRs were, in effect, " voided." The alleger had personally prepared only a few (unspecified) NCRs, and was not alleging that "many" NCRs were voided. Only one specific instance (NCR C82-00060) was identified by the alleger. The alleger considered this NCR to have been " voided" because, instead of sandblasting and entirely recoating hangers and shims, the disposition of the NCR permitted localized repair to be made. The alleger believed this disposition of the NCR was contrary to both site procedures and ANSI standards. Comanche Peak SS'ER 9 M-108
The TRT located NCR C82-00060, Revisions 0 and 1, in the site record vault. This NCR involved a deposit of oil-based soot from forced curing heaters on 16 electrical hangers and 51 hanger shims. This deposit had occurred while the final coat was still " tacky." The NCR dispositioned the hangers for rework by requiring that the contaminated areas be wiped with solvent and, if necessary, lightly sanded until the discoloration was removed. After repair, these areas were rechecked for film thickness. Because of the small amount of coated surface which would be exposed after installation of the shims, the shims were dispositioned "use-as-is." The TRT found the dispositioned repair to be technically adequate and consist-ent both with the site procedures referenced in the NCR and with ANSI standards.
- 5. Conclusion and Staff Positions: The TRT concludes that these allegations are not substantiated. The overall NCR system is in compliance with TUGC0 procedural requirements. However, the TRT noted weaknesses in the Coatings NCR system. These weaknesses are related to insufficient documentation of engineering justifications for those NCRs which involve DBA qualification and coating traceability and to poor trending capability. These concerns
- are assessed in Coatings Categories 2 and 3.
Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original
; concerns and to obtain any additional comments from them. A summary of the followup interviews is included in Section 2.2.3 of this Appendix.
- 6. Actions Required: None.
i i l 3 l i Comanche Peak SSER 9 M-109
Protective Coatings Sc, Design Change Authorizations
- 1. A11eaation Category:
Allegation Number: AQ0-11, AQ0-24, AQ0-26, AQO-27, AQO-28, AQ0-29, AQ0-30 2. and AQO-31 Characterization: It is alleged that: 3. A design change authorization (DCA) allows primer coat of 0.5 mils without design basis accident (DBA) qualification (AQO-11). Protective coatings are placed over rusty, scaly, unprepared metal surfaces inside pipe supports (AQ0-24).
- DCAs are not controlled (AQ0-26).
DCAs are originated by Engineering without QA/QC input (AQ0-27).
- DCAs are written instead of NCRs (AQO-28).
- DCAs are written to overcome problems (AQ0-29).
DCAs are used to downgrade surface preparation and specification AS-31 (AQ0-30 and AQ0-31). In assessing these allegations, the
- 4. Assessment of Safety Significance:
NRC Technical Review Team (TRT) reviewed the DCA system for coatings generically, as described in site procedure CP-EP-4.0 and 4.6, " Design Control and Field Design Change Control," and discusse DCAs are generated The TRT found that the DCA system is complex. The (TNE). the need for immediate changes, critical to construction, arise. when changes are authorized by specified engineering personnel fol two-stage process. approves a DCA onsite. (This authorized person may or may not be a Gibbs In the second stage, the G&H Architect At Engineer this
& Hill [G&H] engineer.)
performs a final review of the design change and approves it. stage, the DCA system is in compliance with ANSI N45.2.11. The TUEC Design Change Tracking Group (DCTG) tr The TRT found that the DCA system Sectiondescribed inProcess." 4, " Design the site procedures com-However, the plies with ANSI N45.2.11-1974, TRT observed failures to fully implement the procedural system in changes to protective coating requirements which affected DBA coatings. tion Tests, Coatings Category 2.) DCA Allows Primer Coat of 0.5 Mils Without DBA Qualificatio The allegation is that101.2-1972.
~
DCA 18,499 allows a primer t qualified in accordance with ANSI The TRT review of DCA 18,489 Revisions 0 and 1, indicatedAl- that TUEC is aware that this 0.5-mil primer thickness has not be M-111 Comanche Peak SSER 9
the issuance of Revision 1, TUEC has listed the unacceptable areas where a 0.5-mil-thick coating of primer was applied in the coatings exempt log (CEL) as Items 8-18. Protective Q Coatinos Placed Over Rusty, Scaly, Unprepared Metal Surfaces (AQO-24). The allegation is that Q coatings (that is, coatings in safety-related areas) have been placed over rusty, scaly, unprepared metal surfaces inside pipe supports made of tube steel which have no end-caps. In these cases, the protective coating gets on the rust inside of the tube. The concern is that this coating material could later crack, scale, come off the inside of the pipe, and then travel to the sumps. The TRT walked through the Unit 1 Containment Building and observed that the structural tube steel supports have protective coatings applied from 0 inches to approximately 3 inches inside the tubing. Most of the areas observed appeared to be coated with a fine mist, apparently from overspray inside the steel tube. The areas in question are very small, and in many cases there is no coating inside the tube steel. It was obvious to the TRT that these areas would be very difficult to clean and prepare for coating. Moreover, there is no regulatory requirement nor TUEC procedural requirement that end-caps be placed on tube steel. The TRT verified that TUEC entered 6,000 square feet of surface in the CEL, which the TRT believes is a conservative estimate, as this amount is only 1 percent of the total coated surface. . DCAs Not Controlled (AQ0-26). The allegation is that DCA documents are not controlled. A TRT review of this allegation indicates that procedures CP-EP-4.6, CP-EP-4.7, and DCP-3 control field design changes, such as DCAs. Comanche Peak Project Engineering (CPPE) initiates the DCA through proce-dure CP-EP-4.6. The original copy _is sent to the Automated Records Manage-ment System (ARMS), which is a part of the Document Control Center (DCC). DCC receives DCAs, logs them, and issues them control numbers. Controlled DCAs for coatings were issued from document satellite stations in the field. These stations assured that craft personnel maintained current revisions to all design changes. The TRT learned that, prior to April 1984, the TUEC coatings QA/QC group had a control box at DCC into which all DCAs affecting coatings were placed. These were controlled copies. The TRT discussion with a certified file clerk for the coatings QA/QC group indicated that the clerk collected all DCAs from the coatings control box at DCC and attached them to the coatings specifications in the field. The clerk was also responsible for making the QC supervisor aware of each DCA. The supervisor was then. responsible for informing QC inspectors of the DCAs. Procedure DCP-3 describes the document control activities at CPSES. Comanche Peak Project Engineering (CPPE) forwarded information copies to the various disciplines designated on a standard distribution list, such as Quality Engineering and Civil Engineering. These copies were not controlled; however, they were not used for construction and inspection. The TRT determined, therefore, that the DCAs at CPSES were controlled procedurally. Comanche Peak SSER 9 M-112 9
TRT document reviews and discussions with QA/QC personnel concern allegation indicated that prior to current procedural revisions t and un-written practices (such as QA/QC group meetings), QC inspectors did no Although the QA distribution list for DCAs has always receive all DCAs. QA decided whether or not the QC inspectors ne The TRT dures or quality, many times QC inspectors DCA. in would not was told, for example, that many of these OCAs concerned coatingsThe TRT fo inaccessible areas being placed in the CEL. case in its examination of these DCAs. If DCAs which relax coating requirements are not distributed Theto QC inspectors, their inspections could be in the conservative direction. TRT found that none of the OCAs they examined made inspection requir more stringent. The allega-DCAs Originated by Engineering; QA/QC Has No Input (AQ0-27). Engineer-tion is that DCAs at CPSES are originated and tota The TRT review of the present program indicated that t d byQA/QC Engineer-has no proce-
. dural requirements for input in the review of DCAs genera eCoating QA ing; however, QA/QC is on the distribution list for TheDCAs.
TRT's observa-Engineers (QE) receive all DCAs related to coatings. tions and discussions with Quality Engineering, Engineering, and QC per-sonnel indicated that all disciplines are aware of a DCA before it reach The TRT review also indicated that it is the place or activity affected.an unwritten policy that QE then Engineering perso indicates DCAs with Quality Engineering before they are issued. co There is no regulatory requirement that Quality Assurance must review or Criterion III, Appendix B, of 10 CFR 50 does approve design changes. require design changes to be reviewed by the original design orga The TRT determined that the present method for controlling DCAs at C violates no procedure, nor Theispresent theresystem a requirement that appears to be commits working TUEC to Ain QA/QC provide input to DCAs.that QA/QC personnel do contribute to This allegation is that DCAs ih DCAs Are Written Instead of NCRs (AQO-28)_.are used " freq a nonconformance report (NCR) should be written. that 40% of the DCAs were for NCR conditions. The TRT randomly sampled 70 DCAs attached to specification ding AS-31, t majority of which were originated for clarification purposes NCRs are to be originated when a or regar conditions in inaccessible areas. deficiency renders the quality of an item unacceptable It or isindeterminate Of the 70 DCAs reviewed, 5 appeared to indicate an NCR condition. ld the TRT opinion that these DCAs indicated indeterminate q have been addressed in NCRs.The time span between the five DCAs indicate this thatcondition.) DCAs were not frequently used for NCR conditions; thus, it appears M-113 Comanche Peak SSER 9
that these DCAs were isolated cases. The TRT review of the descriptions and resolutions of the five DCAs did not indicate any technical concerns regarding the quality of coatings. The TRT determined that DCAs were used as a tool to resolve problems in a timely fashion as they arose and that a few DCAs were written for con-ditions that should have required an NCR. However, the TRT did not find that DCAs were used " frequently and conveniently" to cover up problems or that 40 percent of the DCAs written were for NCR conditions. DCAs Are Written to Overcome Problems (AQ0-29). The allegation is that DCAs were written to overcome a problem which would take considerable time to repair. In other words, DCAs were used to facilitate the completion of a job even though it meant that accepted QA/QC site procedures wculd not be followed. The TRT review of this allegation indicates that DCAs were generated when changes critical to construction arose. An example would be DCA 16,106, Rev. 1. (Refer to allegation AQO-24 in this category.) The TRT noted that it is acceptable industry practice to modify specification require-ments if they cannot be satisfied for one reason or another, provided that basic design criteria are met and safety is not impaired. Downgrading of Surface Preparation (AQ0-30 and AQ0-31). Allegation AQ0-30 is that, on numerous occasions, DCAs were issued to downgrade the surface preparation from an SP-10 to an SP-6 standard preparation; DCAs were written to downgrade specification AS-31 requirements in the Containment Building to AS-30, which is the nonsafety specification. AQ0-31 alleged that QC management interpreted an SP-6 on a DCA to mean "do the best you can"; when difficult access areas were involved, QC management allegedly stated to QC inspectors, "if you cannot get to an area do not worry about it." The majority of DCAs reviewed by the TRT applied to inaccessible areas.
~
The TRT review of DCAs for allegation AQ0-30 indicated that from time to time Engineering did change or downgrade surface preparation from SP-10 to SP-6. Allegation AQ0-31 is similar to AQ0-30. The TRT review and discussions with QA/QC personnel did not substantiate the allegation that QC manage-ment stated to QC inspectors: "if you cannot get to an area do not worry about it." QC inspectors did have questions about DCA 13,140, Revision 1, involving which areas were considered to be inaccessible and what should be QC inspected. DCA 13,140, Revision 2, clarified this issue by including definitive criteria for determining whether an area was to be considered inaccessible. The TRT determined that downgrading specification requirements for inacces-sible areas to a "best-effort," or "use-as-is" standard is accepted indus-try practice. The TRT review of the CEL indicated that virtually all of the inaccessible areas documented by DCAs had been placed in the exempt log. The TRT also determined that, based on exempt log estimates, the total area downgraded is not over 3 percent of the total coated surface. Comanche Peak SSER 9 M-114
i i i i
- 5. Conclusion and Staff Positions: The TRT concludes that all of the allega-tions reviewed were substantiated. These deficiencies, although not of '
safety significance in the coatings area, must be considered in evaluating
- the effectiveness of TUEC's overall QA/QC program. . !
- Following completion of its onsite work, the TRT Coatings Group attempted i i to contact all.of the allegers to discuss its findings of their original I concerns and to obtain any additional comments from them. A summary of
! the followup interviews is included in Section 2.2.3 of this Appendix. , i, '
- 6. Actions Required: None.
i I e t I I i i 4 i i t k Comanche Peak SSER 9. M-115
- 1. Allegation Category: Protective Coatings 6, Coatings Exempt Log (CEL)
- 2. Allegation Numbers: None specifically for the CEL.
- 3. Characterization: In its assessment of allegations concerning deficiencies in protective coatings work, the NRC Technical Review Team (TRT) found that the number and area of the coated items entered into the CEL was a convenient measure of plant coatings with unacceptable or indeter-minate quality. The TRT also found that many items described in the protec-tive coatings allegations were entered into the CEL, and that the area entered in the CEL was a significant fraction of the total area coated.
The TRT, therefore, conducted a ]cneric review of the operation of the CEL.
- 4. Assessment of Safety Significance: The NRC staff has concurred (see Appendix L) with a study by Texas Utilities Electric Company (TUEC) indicating that failure of the coatings inside the Containment Building would not unacceptably degrade the performance of post-accident fluid systems. Thus, qualification is not required for coatings at CPSES, and deficiencies which might result in coating failure do not have safety significance. As a result, the quantity of coatings represented in the CEL also does not have direct safety significance.
However, based on TUEC's prior Final Safety Analysis Report (FSAR) commit-ment to provide qualified coatings inside the Containment Building, coat-ings applied before issuance of Appendix L were required to have been applied as qualified coatings. TUEC's failure to fulfill that prior commitment, as described in this Appendix, indicates deficiencies in the Coatings QA/QC program. These deficiencies, although not of safety significance in the coatings area must be considered in evaluating the effectiveness of TUEC's overall QA/QC program. The CEL was established by TUEC Procedure CP-EP-16.4, " Protective Coatings Exemption Log," to " provide the method for maintaining identification of items and/or areas that do not meet project coating requirements." In conducting its review, the TRT examined this procedure, interviewed several TUEC civil engineering and QA personnel regarding CEL operation and examined a number of documents by which some of the larger areas (1000 square feet or greater) were placed in the log. These documents included TUEC memoran-dum QTQ-416; design change authorization (DCA) 17,142; nonconformance report (NCR) C-84-00710; DCA 6114, Rev. 1; DCA 12,374, Rev. 1; DCA 16,106; and NCR C-84-01488, Rev. 4. The TRT found several deficiencies in TUEC Procedure CP-EP-16.4. Although it places responsibility for approving items to be included in the CEL with a civil engineer or his representative, it provides no specific direction or criteria to assure that items not meeting project coating requirements and not scheduled for repair or rework are systematically entered into the CEL. Nor does the procedure require each CEL entry to be signed and dated by the civil engineering representative, to identify the document describing the coating deficiency, or to describe the basis for _ placing the item in the log. The TRT exanined the documents (NCRs and DCAs) by which some of the larger items were entered into the CEL, which was provided to the TRT by TUEC Comanche Peak SSER 9 M-117 l
letter of August 10, 1984. Also, the TRT examined the method by which the areas designated for the log were estimated. The TRT found that determina-tions to place items in the CEL were made in a conservative manner and the methods for estimating the areas involved were reasonably conservative. In the course of its review of the backfit test program and other aspects of the protective coatings, the TRT found a number of items that should have been included in the CEL, but were not. The largest of these were the areas of miscellaneous steel and concrete and containment liner which failed the coating backfit test program adhesion tests after the original data were corrected for the Elcometer calibration error. This total area may be as large as 57,500 square feet. Approximately 3000 square feet have already been entered into the CEL because of failed adhesion tests; the remaining 54,500 square feet should be added to the CEL. (See Coatings Category 1.) A second item not included in the CEL involved coated areas with deficien-cies other than poor adhesion. For example, a number of NCRs relating to unsatisfactory dry film thickness (DFT) (C-83-03103, Rev. 2; C-83-03104, Rev. 2; and C-83-3105, Rev. 2) direct that all miscellaneous steel items with unsatisfactory DFTs be "used-as-is" and entered into the CEL. As a
- conservative estimate of the area involved, TUEC used 5 percent of the total surface area of each category of miscellaneous steel for a total of 8,150 square feet. This estimate may be low because the DFT test failure rate before cosmetic rework averaged 8.5 percent for the miscellaneous steel categories. Also, according to interviews between the TRT and TUEC QA personnel, a number of discrepant areas were still not finally disposi-tioned by either rework or entry into the CEL.
A third item which has not been included in the CEL involves non-standard coatings which were not DBA qualified. (See Coatings Category 2.) One example is inorganic zinc coatings applied over organic topcoat in the overlap areas that surround repairs to protective coatings over steel. By letter of August 21, 1984, to NRC, TUEC estimated this overlap area to be between 2,500 and 6,500 square feet. Available documents do not indicate whether the 6,100 square feet of inaccessible or limited-access areas described in TUEC's letter TXX-4262 to NRC of August 21, 1984, in relation to allegation AQ0-31 were entered into the CEL. DCA 13140, Rev. 2, downgrades the requirements for these areas, but does not specify that the areas should be piaced in the CEL.
- 5.
Conclusion:
The TRT concludes that the operation of the CEL system was deficient in several respects. TUEC Procedure CP-EP-16.4 did not provide adequate guidance and direction to assure that areas with coatings of unacceptable or indeterminate q'ality u were either reworked or' entered into the CEL. The TRT further concludes that several sizable areas with coatings of indeterminate quality have not been included in the CEL. These areas include 54,500 square feet, which may have failed the adhesion test accord-ing to the TRT audit of the BTP, and 2,500 to 6,500 square feet of non-standard coatings which were not DBA qualified. Therefore, approximately Comanche Peak SSER 9 M-118 s , _ . . . _
000 square feet of area of indeterminate quality should be added to the Without including this additional 60,000 square feet, the CEL identified approximately 55,000 square feet of unqualified or indeterminate coatings This 55,000 square-foot value is already considered high by the, TRT and it would feet. Thebe more total than doubled by including the additional 60,000 square of 115,000 total coated area in the Unit 1 Containment Building. square feet is approx this value is excessive when compared to CEL areas reported by otherThe TRT fin applicants. The implication of the 20 percent CEL value is that the remaining 80 per-cent of the coatings are of satisfactory quality. cation cannot be considered valid until the resolution of other TRT con-However, ability, is reached.cerns, such as assurance of DBA qualification of coatings and their (See Coatings Categories 2 and 3.)
- 6. Actions Required:
TUEC shall provide updated estimates of the additional items, including those detailed above, to be entered into the exempt log Although all coatings are now considered not safety related, the CEL shall . the requirements in effect at the time the coating work w Thiswith ent logthe will be usedofinAppendix guidelines planningL. future inspections of coatings consist-Comanche Peak SSER 9 M-119
- 1. Allegation Category: Protective Coatings 7, Training and Qualification of Coatings Inspectors and Painters
- 2. Allegation Numbers: AQ0-22, AQ0-32, AQ0-33, AQ0-35, and AQ0-61
- 3. Characterization: It is alleged that:
Coatings inspectors must perform backfit test program adhesion testing without first completing training (AQ0-22). Reading list coatents have been changed after inspectors sign the list (AQO-32). A lead coatings inspecto" lacked the qualifications to properly perform his duties (AQ0-33). Some persons providing training to prosprctive inspectors are not properly qualified (AQ0-61a). Level II inspectors " sign-off" for training conducted by Level I inspectors (AQ0-61b). Management was aware that some inspectors were trained by unqualified instructors and took no corrective action (AQ0-61c). Workmanship is poor because painters lack the qualifications necessary to produce quality work, and painter certification documentation is deficient (AQ0-35).
- 4. Assessment of Safety Significance: In assessing the allegations, the NRC Technical Review Team (TRT) reviewed the applicable training and qualifi-cation requirements. In the CPSES Final Safety Analysis Report (FSAR),
Texas Utilities Electric Company (TVEC) commits to American National Standards Institute (ANSI) standard N101.4-1972, " Quality Assurance for Protective Coatings Applied to Nuclear Facilities," and Regulatory Guide 1.54 (June 1973), which endorses ANSI N101.4-1972, with comments. ANSI N101.4-1972 requires that the QA program include provisions for the qualification of application and inspection personnel. Effective with Revision 15 (April 30, 1981) of the FSAR, TUEC committed to NRC Regulatory Guide 1.58, Revision 1, with minor modifications. This regulatory guide endorses, with comments, ANSI N45.2.6-1978, "Qualifi-cation of Inspection, Examination, and Testing Personnel for Nuclear Power Plants." ANSI N45.2.6-1978 provides guidelines and criteria for the evaluation and qualification of inspection personnel. CPSES specification 2323-AS-31 " Protective Coating," in Appendix C, para-graphs 6.3a and 5.2bl, respectively, requires certification of coatings inspectors and applicators. Requirements for Certification of Inspection Personnel. The TRT reviewed each revision of the Texas Utilities Generating Company (TUGCO) procedures providing methods and requirements to qualify coatings inspectors. These procedures are: CP-QP-2.1, " Training of Inspection Personnel," CP-QP-2.3,
" Documentation Within QA/QC Personnel Qualification File," and QI-QP-2.1-4, Comanche Peak SSER 9 M-121 1
" Qualification of Protective Coating Inspection Personnel." These pro-cedures also specify requirements for documentation of training, testing, and certification activities.
ANSI N45.2.6-1978 defines the minimum capabilities required for each level of inspector certification. Capability is to be established by suitable evaluations of education, experience, training test results, and capa-bility demonstration. The TRT examined evidence of capability evaluations, as discussed below. The TRT reviewed the complete qualification files of 23 QC personnel certified between 1978 and October 1984, and portions of other files described hereafter. The TRT also interviewed 11 Level I and Level II inspectors, 2 Level III Quality Engineers (QEs), 2 former QC supervi-sors, and the former QA manager. The TRT did not contact coatings person-nel not presently on site relative to training and qualification. Certification of Level I Inspectors. The TRT found that the previous experience of inspectors was not in every case evaluated in accordance with applicable requirements. ANSI N45.2.6-1978 and TUGC0 procedure CP-QP-2.1 require certain "related experience in equivalent inspection, examination, or testing activities" for each level of certification. ANSI N101.4-1972 states that an inspector's " qualifications shall include his prior training and inspection experience for work of comparable scope with generic coating systems similar to those used for the work in question." The TRT considers that this statement defines " equivalent" experience when cited as a qualification basis. The TRT found previous experience as a journeyman applicator (painter) credited on inspector certifications, including the certification of four presently employed inspectors, as a basis for qualification. The TRT con-siders that, although experience as an applicator of nuclear coatings is beneficial, it does not constitute equivalent inspection experience, and is not a suitable basis for qualification, nor a suitable basis to waive indoctrination and training requirements. The TRT found that education and previous work experience used as a basis of inspector qualification were not adequately documented in all cases. The TRT found 14 instances where the verification of education and previous experience was missing or incomplete.- In a few of these cases, such verification was available from other uncontrolled files. Factors other than education and experience may demonstrate inspector capability. ANSI N45.2.6-1978 and TUGC0 procedure CP-QP-2.1 provide that required education and experience levels should be interpreted to recognize that job performance, training, and testing may provide reasonable assur-ance that an inspector may competently perform a task. In such cases, Regulatory Guide 1.58 requires documentation that demonstrates that the inspectors have the competence that would-have been gained from the required education and experience. The TRT also evaluated other available evidence to demonstrate each inspector's capability. ANSI N45.2.6-1978, requires that on-the-job training (0JT) be included in the training pro-gram. The TRT questioned inspectors concerning 0JT which they have given or received and found that their training regularly includes 0JT, which Comanche Peak SSER 9 M-122
consists of inspector trainees performing actual inspections. The TRT found that OJT serves as a primary training vehicle for personnel without equivalent previous experience, which is one acceptable method for meeting experience requirements. 0JT records examined by the TRT showed the procedure (s) covered during the OJT session, the date and time they were covered, and the names of the trainee and instructor. However, these records did not provide suf ficient detail to assess the quantity and quality of the 0JT. OJT records did not identify activities, functions, or inspections performed, did not show that acceptable inspection documentation was prepared, and did not show that the requirements of the inspection work were successfully demonstrated by the trainee. The TRT found that 0JT requirements were often waived on the basis of pre-vious experience and/or examination results. For example, OJT requirements were partially or completely waived for 14 of 24 inspectors certified to perform backfit test program inspections. Further, the TRT found that all examinations did not include practical tests for important operations, such as visual identification of film defects. In its review of selected Level I examinations, the TRT found that written examinations primarily test the inspector's ability to identify specific work and inspection criteria, such as film thickness and ambient condition requirements. The TRT found that written examinations did not test the inspector's knowledge of applicable QA requirements, such as for instrument and document control, the method of operating instruments or performing important inspection tasks, or-the precision characteristics of inspection methods and instruments. The TRT reviewed selected examples of the answer keys used to grade examinations, and compared the responses given in answer keys to those given on examinations. The TRT found inconsistently or improperly graded examinations. In one instance, the answer key gave the answer as a description of the method for an inspection, whereas a test response that identified only the name of the inspection was given full credit. The TRT considers that these inconsistencies may obscure the inspector's inability to understand important requirements. A number of examinations consisted of written and practical (e.g., demon-strated ability) test elements. On the basis of interviews and a review of records, the TRT learned that practical tests were graded on a pass / fail basis without written guidelines for grading Practical tests included in various examinations provided 24 percent to 50 percent of the grade of the combined written and practical examination. The TRT noted that, as a result of scoring 50 points on the practical test elements, inspectors who achieved the required passing grade of 80 percent had, in fact, scored only 60 percent on the written portion of the examination. The TRT considers that the formal classroom training did not assure the capability of inspectors because the training was primarily a review in preparation for the corresponding examination, and did not include a lesson plan showing an adequate review of necessary inspection requirements. Comanche Peak SSER 9 M-123 1
In one instance, a presently employed inspector was shown to have a color vision deficiency. The type of deficiency was not noted, and there was no objective evidence of the inspector's visual acuity under field conditions. One of the inspector's annual examinations, during which ne failed to identify 9 of 12 test plates, was waived on the basis of a supplementary test given with colored pens. The TRT finds that documentation provided by Level I int?ector files was not adequate or sufficient to demonstrate that the' required capability of all personnel had been achieved. The TRT randomly selected 20 inspection reports generated between 1978 and October 1984, to verify the qualification of each inspector to perform the documented inspections at the date of the report. The TRT found no records for the inspector who signed report PC03511, of February 24, 1979. The signature is difficult to read, and the TRT could find no record of a similar signature for this individual in the Permanent Plant Records Vault, although the TRT found certification documentation for the other inspectors. The TRT evaluated the process of Level I inspector recertification to determine how the proficiency of inspectors was maintained. The TRT reviewed approximately eight completed recertification forms, which showed that written, oral, or practical recertification examinations had been given. However, the TRT found no supporting evidence that written, oral, or practical examinations had actually been given for inspector recertifi-cation. The TRT found that recertification consists of an informal evaluation by the responsible QC supervisor, based on personal . knowledge and supplemented by information from the responsible lead inspector. TRT interviews disclosed that to maintain their proficiency, inspectors are directed to read revisions to governing documents. The TRT found documen-tation for one instance of required supplemental reading included in several inspectors' files, in accordance with the requirements of TUGC0 procedure CP-QP-2.1. The TRT found no verification of reading, or records of formal training for other important revisions to governing documents, such as revision 2 of CPSES specification 2323-AS 31. ;
)
The TRT considers that the methods used at CPSES to recertify Level.I coatings inspectors do not demonstrate the continued capability of these personnel. During the course of its interviews, the TRT asked inspectors j to explain a number of generic and specific requirements and methods of implementing those requirements. These discussions were not detailed enough to fully assess the inspectors' understanding of requirements; however, in the opinion of the TRT, each inspector demonstrated an ade-quate understanding of methods and requirements for the topics covered. i Certification of 0JT Instructors. TUGC0 Procedure CP-QP-2.1, beginning with Revision 8, requires that OJT be conducted under the direct super-vision of someone (e.g. , Level I) certified as an OJT Instructor, or a Level II or Level III inspector. The basis of qualification of 0JT Instructors is stated on the certification form given as an attachment to the procedure. The TRT interviewed coatings inspectors, and coatings QA Comanche Peak SSER 9 M-124
Review of Indiv_idual Allegations Backfit Test Program Training (AQO-22). The allegation is that coatings inspectors are required to conduct adhesion tests of applied coatings, as part of the backfit test program, without first completing formal training in the proper methods to be used in performing the tests. The testing method cited by the allegation requires operation of the Elcometer Model 106 adhesion tester. The implied significance of this allegation is that the tests may have been improperly performed and that the data generated may have resulted in the improper acceptance of coatings. (A discussion of the history and requirements of the backfit test program and the evalua-tion of the backfit test program inspection data are given in Coatings Category 1.) Coatings inspectors performing backfit test program inspections are certified to TUGC0 procedures QI-QP-11.4-23, " Reinspection of Seal Coated and Finish Coated Steel Substrates for Which Documentation is Missing or Discrepant," and/or QI-QP-11.4-23, " Reinspection of Protective Coatings on Concrete Substrates for Which Documentat'on is Missing or Discrepant." Each proce-dure requires the use of the Elcometer adhesion tester. The TRT randomly selected 20 backfit test program inspection reports, and found that the inspectors who prepared the reports were properly certified at the date of each report. The TRT's review of logs since 1982 found one NCR, C-83-00852, which dispositioned 19 backfit test program inspections performed by an uncertified inspector. The TRT also examined the files of the 24 QC inspectors certified to TUGC0 procedures QI-QP-11.4-23 and/or QI-QP-11.4-24 between the start of the backfit test program and the present. As previously mentioned, ANSI N45.2.6-1978 requires that training include 0JT. This standard further provides that uncertified personnel may be used "in data-taking assignments" provided they are supervised by qualified personnel. 0JT for backfit test program adhesion testing and Tooke gauge film thickness measurements generally includes data-taking operations. Regulatory Guide 1.58 requires that, in such an instance, these trainees have completed training adequate to assure competent performance prior to performing data-taking 0JT. During its review of files, the TRT noted ten instances where records show that OJT for backfit test program training preceded study of appropriate requirements. The TRT also noted seven instances where 0JT for backfit test program certification was completely waived; however, in each case the inspectors studied requirements and completed a written / practical exam prior to certification. The TRT found that inspectors have been certified for this inspection on the basis of the single demonstration of capability during a practical examination. The TRT has commented above on practical examination grading practices. The TRT found that the training program does not require that an inspector trainee-complete a formal training session or read applicable instructions prior to performing inspection tasks while participating in 0JT. During interviews with inspectors, the TRT found that trainees are instructed to read the applicable requirements prior to 0JT, but there is no consistent verification that the instruction is followed. Comanche Peak SSER 9 M-127 l
performance and capabilities of other inspectors; and, contacting respon-sible QEs on questions or problems beyond the capability of the lead inspector. The TRT concluded that the position is verbally defined and assigned by a responsible supervisor. The TRT found that TUGC0 procedures require that certain functions be performed by lead inspectors. For example, TUGC0 procedure QI-QP-11.4-26, Revision 6, paragraph 2.3.1.1.c, assigns lead inspectors to the task of reviewing and approving a log of environmental inspections. Further, certain requirements documented on Protective Coatings Technical Training Outlines are verified and signed by a lead inspector, and the responsible supervisor may indicate completion of other items on these outlines on the basis of statements by lead inspectors. The TRT interviewed six recently certified Level II inspectors to determine their scope of training and examination. The Level II certification pro-cess consists essentially of an undocumented study and review of current requirements, a documented classroom review of requirements, and a written examination. The TRT reviewed approximately six Level II examinations. The examinations tested knowledge of specific objective requirements in greater detail than for Level I examinations. The examinations did not demonstrate or estab-lish the individual's capability either to perform the functions described by ANSI N45.2.6-1978 as listed above, or to perform the functions regularly required of lead inspectors and Level II inspectors in the coatings inspection program. Certification of Level III Coatings Personnel. The TRT found that indi-viduals assigned as QC supervisors at CPSES typically have Level.III certifications. TUGC0 Procedure CP-QP-2.1, Revision 16, provides that Level III Civil certifications include protective coatings in their area of responsibility. The TRT found one recent supervisor in the coatings QC inspection department who was certified as a Civil Level III, whose file does not include documented study of coating _ requirements and procedures, and whose resume does not clearly demonstrate previous technical experience with the generic coatings systems used at CPSES. As a supervisor, this individual attested to the capability of a number of inspectors and made other technical decisions concerning coatings inspection work. TUGC0 letter TXX-4262, August 21, 1984, identifies individuals who have been assigned to supervisory functions for the coatings quality control program since January 1982. It also identifies three individuals assigned such functions who did not have certifications for coating work. The TRT found that these individuals have signed QC inspector certification docu-mentation, as QC supervisors, to verify evaluations of capability, and have made other technical decisions concerning coatings inspection work. TUEC records do not demonstrate the capability of QC supervisors to l perform their assigned functions. The TRT was unable to assess the effect of this finding upon the quality of coatings applied at CPSES. Comanche Peak SSER 9 M-126
accurate. As noted previously, the TRT found that evidence of compliance with the periodic reading requirements given by TUGC0 procedure CP-QP-2.1 is not generally present in inspector personnel files. Qualifications of Lead Coatings Inspectors _(AQ0-33). The allegation is that many problems with the coatings inspection program occurred because of the inexperience of one of the lead inspectors in the coatings QA/QC department. The allegation cites an instance in which this individual identified rust on an A-frame at the Seal Table room as being residue of Ameron D-6 primer. The implied significance of this allegation is that inadequate technical knowledge by supervisory personnel will contribute to defective inspection work and to defective analysis of the results of inspection. In assessing this allegation, the TRT notes that the various oxides of iron are typically described as " reddish-brown" to " black"; the color of the primer coating is typically described as " reddish grey." The TRT reviewed the permanent file of the lead inspector named in the allegation. The inspector's resume shows 5 years of experience as a journeyman and foreman painter for nuclear coating work, and 19 years as a journeyman painter for industrial and commercial coating work. Previous coatings inspection experience was limited to one 3-month period of intermittent coatings inspection. The TRT found no indication on the individual's resume of training in the technical characteristics of protective coating systems or in coatings inspection methods prior to this person's arrival on site. Verification of previous employment and education was not in the file, but experience and education were cited as a basis for qualification. Verifi-cation efforts by TUEC during the course of the TRT review showed that the individual had not received a high school diploma as claimed, and that the work experience claimed was inaccurate. The TRT found that certifications for this individual completely or partially waived OJT requirements on the basis of his previous experience. The TRT also found that answers on this person's qualifying examinations differed from the answers on the prepared answer key for the examinations. The TRT found that this person was certified to perform backfit test program inspections 2 weeks after being hired, and as a Level I inspector for other functions 4 weeks after hiring, was also appointed as a lead inspec-tor within 4 weeks of hiring, and was certified as an 0JT Instructor approximately 6 weeks after hiring. As a lead inspector, this individual verified that inspector trainees had successfully completed requisite training. (The individual named by the allegation was released from employment in about May 1983.) The TRT reviewed documentation of the incident cited in the allegation, including inspection reports PC47854 (11/4/82), PC47874 (11/5/82), and NCR C-82-02403 (12/30/82), with attached inspection reports. The TRT i found evidence in these records to demonstrate that three other inspectors believed the residue was rust, although there was evidence that respon-sible QEs disagreed with the evaluation made by these inspectors. Comanche Peak SSER 9 M-129
i The TRT noted that TUGC0 procedures QI-QP-11.4-23 and QI-QP-11.4-24 do not provide a detailed operational method for operating the Elcometer 106 adhesion tester. CPSES specification 2323-AS-31 directs that the instrument be operated in accordance with the manufacturer's recommenda-tions. The TRT found that the Protective Coatings Technical Training Outlines do not show that inspectors have read the manufacturer's recommendations. The TRT could find no copy of the referenced manufac-turer's instructions available at the QC inspectors' office or in the possession of inspectors, although a copy of the instructions was avail-able from responsible coatings QEs upon request. Since about July, 1984, TUGC0 procedure QI-QP-11.4-29, "Use of Elcometer Adhesion Tester for Isolating Areas of Questionable Coating Adhesion," has been employed, as stated by the title of the procedure. This procedure describes a detailed operational method for use of the instrument. The TRT noted a separate but similar deficiency in that TUGC0 procedures do not provide a detailed operational method for the use of the magnetic dry film thickness gauge used for regular inspection work at CPSES (e.g. the non-backfit test program). CPSES specificatior 2323-AS-31 requires that this inspection be conducted in accordance with a standard consensus specification, Steel Structures Painting Council (SSPC) specification PA2-73T, which provides detailed instructions. The files of three current inspectors show no evidence that they have studied SSPC PA2-73T. In the opinion of the TRT, based on an informal consensus of industry experience, the majority of possible operator-induced errors in the opera-tion of the adhesion tester will result in lowering rather than raising the indicated adhesion values, and thus would show failure of coatings with acceptable adhesion. Further, it is the opinion of the TRT that, although misoperation of the instrument may adversely affect its calibra-tion accuracy, the probability of this occurring is low. (The effect of calibration accuracy on the backfit test program data is assessed in detail in Coatings Category 1.) The TRT considers that this allegation is substantially correct in that the training program does not provide adequate measures to assure that trainees are fully familiar with instructions and requirements prior to OJT; nor does the program demonstrate adequate inspection experience or testing where 0JT was waived. However, such training as was given was beneficial. Further, the probability of the practices described affecting the backfit test program adhesion test data in a non-conservative manner is low. Reading Lists (AQ0-32). The allegation is that after.a reading list was signed by QC inspectors, the document which they had acknowledged reading was exchanged for another document while the reading list acknowledgement form remained the same. The significance of this allegation is that inspectors might be held responsible for implementing requirements of which they are unaware. The TRT did not identify any instance in which a reading list prepared as part of initial training showed evidence of being altered. The TRT reviewed certain reading lists with current inspectors and found the lists-to be Comanche Peak SSER 9 M-128
F-examined painter certification files, and interviewed painters and super-visory personnel. ANSI N101.4-1972 requires that " application personnel shall be qualified in accordance with the coating applicator's qualification procedures." The TRT found that B&R procedure CCP-30, " Coating Steel Substrates Inside Reactor Building & Radiation Areas," CCP-30A, " Coating Steel Substrates Inside Reactor Building & Radiation Areas," and CCP-40 " Protective Coating of Concrete Surfaces" have provided qualification methods in each revision. The painter certification form corresponds to the form recommended by ANSI N101.4-1972. The TRT interviewed two painter foremen and four journeymen painters and verified that the qualification files of these painters correspond to the information given by a current painter qualification summary listing distributed for use in the field. During the interviews, the TRT asked questions to establish the character and extent of the training given to journeymen. The TRT found that training cited by certifications was given as stated. The TRT found that these journeymen exhibited an adequate knowledge of work requirements. During a review of logs since 1982, the TRT noted only one NCR resulting from application by an unqualified journeyman, NCR C-83-00310 (Revision 1, January 31, 1983). The TRT interviewed one painting supervisor who has supervised painter demonstration applications and conducted the formal training session given to painters. The TRT asked questions to establish the nature and. character of training and qualification practices since 1980, and found no discrepancies with the statements made by the painters. This supervisor gave the TRT copies of the lesson plans presently used for classroom training and copies of an examination given to painters upon conclusion of the training session. The TRT reviewed the lesson plans and compared them against project requirements. The TRT found that, in general, the lesson plans contain adequate, useful, and accurate informa-tion that is of substantial value in a well-conducted training session. However, the TRT found the following subjects inaccurately, incompletely, or not addressed by the lesson plans:
- Inspection hold points are not adequately described. The lesson plans do not define " hold point," or include sufficient detail to clearly identify that there are certain inspection verifications required immediately prior to coating application.
- Good application techniques are not fully covered by the lesson plans. There is no evidence of a review of the proper use of spray equipment, such as adjustment of the equipment, techniques to minimize spray application errors, and techniques to spray apply coatings to complex shapes. There is no evidence of discussion of good brushing techniques or the treatment of brushing to accommodate the characteristics of certain materials. (The TRT noted that painters must complete a practical test prior to certification.)
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The TRT concludes that records do not demonstrate the capability of the individual named by the allegation to perform the duties required for his position as Level I inspector, 0JT instructor, and lead inspector. The TRT did not determine the extent to which the inadequate qualifications of this individual may have affected the quality of coatings applied at CPSES. However, the instance cited by this allegation demonstrates an example of defective evaluation on the part of inspectors or responsible QEs. The TRT was not able to clearly identify which group made a defective evaluation, but noted that the adhesion of the coatings on the A-frame failed after the original engineering identification of the residue as primer. Repair of failed areas is documented in NCR C-82-02403. Qualification of Instructors (AQO-61a, AQO-_61b, AQO-61c). These allegations are that Level I inspectors not properly qualified to provide 0JT instruc-tion are training other Level I inspectors, that management was aware that some inspectors were trained by unqualified instructors and took no corrective action, and that Level II inspectors sign-off for training conducted by Level I inspectors. The implied significance of this alle-gation is that inadequate training of new inspectors will result, and that these personnel will not be able to provide adequate inspection of coating work. In assessing this allegation, the TRT reviewed the OJT records for six inspectors certified prior to October 1983, and the files of ten inspectors certified since that date. This review identified 30 individuals who conducted 0JT. The TRT found two instances of Level I inspectors, not yet certified to conduct OJT, who signed 0JT records as instructors during 1984. The TRT found no evidence that this deviation from requirements was detected or identified as a nonconforming condition even though management sign-off of 0JT records and instructor certification was evident. The TRT found no indication that OJT was conducted by individuals other than those who had signed the 0JT record form, or that formal training was conducted by individuals other than those documented on available records. The TRT has noted that lead inspectors and QC Supervisors verify the com-pletion of various training requirements, as indicated by their signatures on the inspector's Protective Coatings Technical Training Outline. Qualification of Painters (AQ0-35). The allegation is that the abilities of coating application personnel are inadequate and that there are problems with workmanship, quality of work, and the indoctrination and qualification of painters. The allegation cites as an instance that the documentation of painter qualifications and in process work did not consistently satisfy requirements. The implied safety significance of this allegation is that unqualified painters might produce defective work. This allegation does not identify areas for which the documentation of in process work is deficient. (The TRT has addressed general and specific concerns about work and inspection procedures and the adequacy of coating work documentation in Coatings Categories 1 through 6.) In assessing this allegation, the TRT examined the requirements, methods, and practices used to qualify painters to apply coatings. The TRT also ! reviewed the requirements of applicable Brown & Root (B&R) procedures, l 1 i Comanche Peak SSER 9 M-130 , i
The TRT concludes that allegations AQO-61a and AQ0-61b are substantiated and that the coatings quality control program lacks sufficient controls to assure the capability of instructors or the accurate assessment of capabil-ity demonstrated during training. The TRT concludes that allegation AQ0-61c cannot be supported by tangible evidence. The TRT concludes that allegation AQ0-35 was partially substantiated by the evidence examined; however, the TRT considers the specific deficiencies noted to be without significance to completed work. The TRT's overall evaluation of the coatings QC inspector qualification system is that, as implemented, it lacks controls sufficient to assure the required competence of personnel performing functions ascribed to Level I, Level II, and Level III personnel by ANSI N45.2.6-1978. The TRT's overall evalu'ation of the painter qualification system is that, with the exception of specific noted deficiencies, the system is adequate to satisfy all requirements. These deficiencies, although not of safety significance in the coatings area must be considered in evaluating the effectiveness of TUEC's overall QA/QC program. Following completion of its onsite work, the TRT Coatings Group attempted to contact all of the allegers to discuss its findings of their original concerns and to obtain any additional comments from them. A summary of the followup interviews is included in Section 2.2.3 of this Appendix.
- 6. Actions Required: None.
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- There is no evidence of training (or testing) in the proper and accurate use of the instruments regularly used by painters to check the accuracy of their work, such as dry film and wet film thickness gauges.
The TRT found that examinations accurately reflect project requirements, are consistently used, accurately graded, and regularly updated to reflect new requirements. The TRT noted that more complex examinations have been given to certain foremen. The TRT randomly selected the names of six journeymen identified as certi-fied applicators by inspection records between 1979 and 1984, and verified that each journeyman was certified for the specified application on the dates shown by inspection records. The TRT also reviewed the certification files of 14 painters, including the 6 interviewed journeymen. The TRT found that, in some instances, painter certification documentation contains misleading or incomplete information which is cited by certifications as a qualification basis. The TRT is concerned that such statements might be used as a basis to assign painters to work beyond their true level of capabilities. For example, some certifications state that, " employee has previous experience as a painter," without identifying the nature or extent of the stated. experience. The TRT found that the previous experience of certain journeymen does not include full-time professional experience or experience applying the generic coating types used at CPSES, or previous experience with related nuclear, industrial, commercial, or architectural coatings. Further, most certifications state that the employee has " experience with the following product types: zines, phenolines, epoxies, latexes, enamels, and thinners," and do not describe the type of experience. The TRT found that such experience was stated in cases when painters did not have experience with these generic coatings prior to being hired at CPSES, and the cited experience consisted cf providing support to and observing certified applicators at CPSES. The TRT found that, with the. exception of the deficiencies noted above, the CPSES system of qualifying applicators is effective and in accordance with requirements and the general practices used for nuclear coating work.
. 5. Conclusions and Staff Positions: The TRT concludes that AQ0-22 is par-tially substantiated and that personnel have performed backfit test program adhesion testing without first receiving all the appropriate I
training in the methods and requirements for this inspection. The TRT further concludes that this ' deficiency tends to cause data errors in a conservative direction, and so should not have a significant adverse impact upon backfit test program data. 4 The TRT concludes that allegation AQO-32 is unsubstantiated. The TRT concludes that allegation AQ0-33 is substantiated, but was not able to assess the significance of the deficiency alleged. Comanche Peak SSER 9 M-132
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