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NUCLEAR REGULATORY COMMISSION
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December 5, 1979
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MEMORANDUM FOR:
B. H. Grier, Director, Region I J. P. O'Reilly, Otractor, Region II J. G. Keppler, Director, Region-III K. V. Seyfrit, Director, Region IV R. H. Engelken, Director, Region V
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Norman C. Moseley, Director Director, Division of Reactor FRO't:
Operations Inspection. Office of Inspection and ' Enforcement
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SUBJECT:
IE BULLETIN NO.: 79-28, POSSIBLE MALFUNCTION OF MAMCO MODEL EA 180 LIMIT SWITCHES AT ELEVATED TEMPERATURES The subject IE Bulletin should be dispatched for action on December"'7,1979, to all pee $r teactor facilities with an Operating License or a Construction permit.
The text d the Bulletin and draf t letter to licensees are enclosed for this purpose.
/Y4 Norman C. Moseley, Director Division of Reactor Operations Inspection Office of Inspection and Enforcement E/osures:
72 Draft Transmittal Letter
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CONTACT:
V. D. Thomas IE 49-28180
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(Draft letter to all power reactor facilities with an operating license or a construction permit)
IE Bulletin No. 79-28 Gentlemen:
Enclosed is IE Bulletin No. 79-28 which requires action by you with regard to your power reactor facility (tes) with an operating license or a construction persit.
Should you have questions regarding this Bulletin or the actions required of you, please contact this office.
Sincerely, Signature (Regional Director)
Enclosures:
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2.
List of Recently Issued IE Bulletins I
3.
Extract of Letter From NANCO dated August 30, 1979 l
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r UNITED STATES SSINS No.: 6820 D
NUCLEAR REGULATORY COMMISSION Accession No.:
OFFICE OF INSPECTION AND ENFORCEMENT
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7 WASHINGTON, D.C.
20555 December 7, 1979 IE Bulletin No. 79-28 POSSIBLE MALFUNCTION OF NAMCO MODEL EA180 LIMIT SWITCHES AT ELEVATED TEMPERATURES Description of Circumstances:
The NRC has been recently advised through a 10 CFR 21 report from NAMCO Controls that a malfunction of a NAMCO Hodel EA180 stem mounted limit switch (SNLS) occurred at the Cooper Nuclear Station.
Investigation into the switch failure by the licensee revealed yellow and brown " crystal-like" resin deposits on the internal components of the switch. The affected switch is located inside the drywell containment at this facility and was being used as the replacement switch for an unqualified SMLS previous 1v identified in IE Bulletin Nos.
78-04 and 79-01.
According to the manufacturer, the problem was traced to a batch of top cover gaskets of which some were over-impregnated and insufficiently heat cured.
It has oeen determined that this condition can leave an uncured residue of " Loctite" in the gasket, which vaporizes at sustained temperatures above 175*F.
To correct the problem, the manufacturer has revised production techniques beginning r
September 1979 in order to better control the impregnation process and to properly heat cure the gaskets following impregnation. This problem is unique to all NAMCO Hodel EA180 series switches received by licensees after March 1, 1979. According to the manufacturer, the suspect switches can be identified by checking the date code which is a 4 digit number stamped on the conduit boss of the switch housing.
NAMCO recommends that any EA180 series awitch with a date code between 02-79 through 08-79 should have its top cover gasket replaced. Also, licensees should request from their suppliers of equipment on which NAMCO EA180 series switches are used that they check their inventory and replace top cover gaskets on switches date coded between 02-79 through 08-79.
1
'he enclosed letter from NAMCO further describes the high-temperature environ-i i
l mental problem with the top cover gaskets used in their EA180 switches and provides recommendations to correct the problem. According to NAMCO, this letter has been sent to each customer who was shipped EA180 switches between February 21, 1979, and August 24, 1979.
Action to be Taken by Licensees of Power Reactor Operating Facilities and Holders of Construction Permits:
Determine if your facility has installed or plans to install NAMCO EA180 1.
switches in any safety-related equipment located inside or outside containment, including valve position indicating circuitry related to containment isolation valves.
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i iE Bulletin No. 79-28 December 7, 1979 Page 2 of 2 J.
2.
If such switches are identified, examine the four digit number stamped on the conduit boss of the switch housing.
If this number falls between 02-79 and 08-79, replace the top gasket of the switch in accordance with the manufacturer's recommendations provided in the enclosed letter.
3.
Submit your plans and programs, including schedules for corrective action, regarding your findings in response to Items 1 and 2 above.
4.
Provide the response in writing within 30 days for facilities holding an operating license and within 60 days for those holders of construction permits. Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the U. S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C.
20555.
Approved by GAO, B180225 (R0072); clearance expires July 31, 1980. Approval was given under a blanket clearance specifically for identified generic problems.
Attachment:
Extract of NAMCO Controls l
Letter dated 8/30/79 l
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NAMC0 C0NTR0LS I
August 30, 1979 Attention:
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Subject:
Possible malfunction of Namco Model EA180 limit
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switches at elevated ambient temperatures Gentlemen:
Namco Controls has determined that the EA180 series limit switches shipped
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may have i
against your purchase order (s) i top cover gaskets which will emit a resin vapor at temperatures above 175'F.
This vapor could condense into deposits on the normally open contacts, j
possibly causing a switch malfunction, l
l We have reported this situation to the NRC, and we are contacting each customer J
who has received switches with gaskets from the questionable production lot, with the following reconnendations:
l If the switches have not been put into service, and the ultimate airbient a) temperatures are unknown, the top cover gasket should be replaced.
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If the switches are in service and are subjected to a continuous ambient
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b) temperature of more than 175'F, they should be inspected for deposits, I
the contacts cleaned if necessary, and have the top cover gasket replaced.
If the switches are in service, (or scheduled for service) in a c) continuous ambient temperature of 175* F or less, no action is necessary.
We will provide you with whatever number of new top cover gaskets you mayor, if require, free of charge.
to our Jefferson plant freight collect, for gasket replacement and/or cleaning.
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We ' sincerely regret any inconvenience this remedy may cause.
If you have any l
216-268-4200 or John R.
questions, please contact either Robert H. Kantner, l
8endokaitis, 216-576-4070.
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IE Bulletin No. 79-28 Enclosure December 7, 1979 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.
4
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79-27 Loss Of Non-Class-1-E 11/30/79 All power reactor Instrumentation and facilities holding Control Power System Bus OLs and to those During Operation nearing licensing 79-26 Boron Loss from BWR 11/20/79 All BWR power reactor i
Control Blades facilities with an
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OL l
79-25 Failures of Westinghouse 11/2/79 All power reactor BFD Relays In Safety-Related facilities with an Systems OL or CP 79-17 Pipe Cracks In Stagnant 10/29/79 All PWR's with an (Rev. 1)
Borated Water System At OL and for information PWR Plants to other power reactors 79-24 Frozen Lines 9/27/79 All power reactor e
i facilities which have either OLs or cps and are in the late stage of construction 79-23 Potential Failure of 9/12/79 All Power Reactor Emergency Diesel Facilities with an Generator Field Operating'Lfcense or Exciter Transfc,rmer a construction permit 79-14 Seismic Analyses For 9/7/79 All Power Reactor (Supplement 2) As-Built Safety-Related Facilities with an Piping Systems OL or a CP 1
79-22 Possible Leakage of Tubes 9/5/79 To Each Licensee of Tritium Gas in Time-who Receives Tubes pieces for Luminosity of Tritium Gas Used in Timepieces for Luminosjty 79-13 Cracking in Feedwater 8/30/79 All Designated (Rev. 1)
System Piping Applicants for Ols 79-02 Pipe Support Base Plate 8/20/79 All power Reactor (Rev. 1)
Designs Using Concrete Facilities with an (Supplement 1) Expansion Anchor Bolts OL or a CP
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TUGCD Inspection end Enforcement Bulletin (IEB) record files were incomplete. The bulletin iiles were dec entr al l r ed and not located in the OA records center.
Furt her, t he engineeri ng evaluations were retained by the individual engineers.
9.
There wer e def icienci es in the procedure to process I E Bs. (Not e: The record is not clear as to vnet the inspector meant by deficeint. It is assumed that the deficiencies were the 1ack of a focal point for IEB's and t h e p er c,if ved i l l e def i ci enci es. )
- 10. No f ocal point at TIJGCO to tr ack IED actions.
DISCUSSION:
The MRC issues Inspection and Enforcement Bulletins (IEB) to operating and construction reactor facilities to transmit Information, request action or request information regarding matters of safety, safeguards or environmental significance.f1J The saf ety inf ormation transmitted can identify eneric equipment or design deficiencies.
Licensee's are expected to determine the applicability of the bulletin to their plant and initiate appropriate corrective actions. IEB's usually require licensees to respond with such information as applicability to the plant, equipment affected, operability status of systems, corrective actions initiated and schedules for completion.
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The NRC inspection program requires that the licem ee's implementing program for IEB requested actions be inspected to assure appropriate actions have been taken.[23 The #cd
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-program requires that all documents in the licensee's response to the NRC be reviewed and a determination be made that the response was proper. It also requires that onsite sampling inspections be made to verify that equipment changes were offected.
From the foregoing, it can be seen that IEB's can be a source for identifying nonconforming conditions and the licensee must make provisions for receiving, evaluating, i ni ti ating correcti ve actions, verifying corrective actions,
& M. reporting results to management and the NRC.
The licensee has a procedure for processing correspondence from the NRC, Nuclear Operations Engineering Manual.
Licensing, Procedure No. NDE-205,includf/g IEB's. This procedt tre as si gns responsi bili ti e 4, and discusses the processing and the retention of ecords.
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SAFETV SIGNIf'ICANCE:
A e+ umi n g the worst cese to be the bred down in the proce ssi n g of IEP s.
the impact on safety is significant. A bu))etin, by definition. Is only issued when saf ety concerns have been identified and the NRC believes there i ts a threat to the public safety. Bul1etins can affect operatiog and tl[ rough and comprehensi ve con %t ruct i on f uncti ons and a
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c on t r o'] program m u ss t be est ab] 1 shed to evaluate, track and rasolve bulletin issues.
FOLLOWUP ACTIONR AND RECOMMENDATIONS The 1icensee's procedure No. NOE-205, dated October 7,
- 1985, gent ally describes the process for controlling bulletins through the receipt, logging in, r evi ew, pl an development, response and c 1 osecut. Paragraph 4.2.15 specifies that documents which provide source information for the bulletin response be included in the document package. If the procedure is implemented it should provide en adequate system f or processing bulletins. However, it does no"{
clearly discuss the interface between the operating f u c c.
orgeniration and the correr -i va actions systems. For WL-tt3 internal audit functions, '
management and the NRC to be
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abl e t o det er mine if appropriate actions have been taken.
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better definition of the interfece is recommended.
The licensee has committed to perform a procedure and record review to ascertain the adequacy of the IEB program. The review should assure that all bulletins were received.
processed and corrective actions initiated through the established programc.
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Inssection Report 50-445/85-16;50-446[85-13 Items 9 and 10 l '.
DIA Statement of the Issue (See Attachment MM)
Issue.
Resolution in Final Report 9.
Deficiency in TUGCo's procedures Violation downgraded to to handle ~IEBs.
unresolved item l
(445/8516-0-02; 446/8513-U-02).
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No focal point at TUGCo to track Violation / unresolved item IEB actions.
dropped.
2.
Expanded Description of the Issue and Related Background Information I
The OIA-statement of this issue is in error.
It should be revised to indicate that the inspector had only verbally stated he believed TUGCo to be in' violation. Both issues were addressed as unresolved items.
The basic issue as understood by Region IV management is as follows:
The inspector (H. Phillips) in his verbal briefing of his supervisor I
believed that, although the TUGCo operations department had i
i Procedure N0E-205 which deals with I'E Bulletins which require action,
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s 30 because TUGCo had no procedure which described the construction organization responsibilities or how the construction organization was to handles IE Bulletins requiring action or established a specific
" construction IE Bulletin focal point coordinator" that TUGCo was in violation.
Initially in draft 3, paragraph 5 (Attachment 4), the inspector identified this issue as unresolved.
In draft 4, paragraph 4, Regional management replaced the inspector's paragraph with a paragraph which indicated that TUGCo had committed to perform a review of related procedures and records to determine adequacy of the procedures and completeness of associated records as an unresolved item.
3.
Safety Significance of the issue i
Regional management concludes that there is no safety significance to r
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l establishing requirements for which there was no regulatory basis.
The Nuclear Operation Engineering Manual, Procedure N0E-205, Revision 1, (October 1985) " Licensing" ( Attachment 15), clearly established I
responsibilities within the TUGCo organization for handling IE Bulletins l
requiring action.
In summary:
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31 The corporate nuclear licensing supervisor is responsible for forwarding all applicable licensing correspondence to the operation support superintendent (055).
b The OSS is responsible for coordinating the review and response.
The technical support engineer (TSE) is responsible for maintaining a Licensing Correspondence Review (LCR) log and assigning a responsible engineer.
j The TSE assigned engineer is responsible for identifying actions required and to develop a plan.
When action is required, assistance is to be provided by station personnel or TUGCo Nuclear Engineering.
(TUGCo Nuclear Engineering is the proper organization to initiate action on the construction side.)
TUGCo Nuclear Engineering would initiate construction activities as required, using existing QA program or construction orogram controls, such as Design Change Authorizations, TUGCo Design Deficiency Reports, Design Calculations, QC Inspection, etc. These programs have governing procedures which outline responsibilities and overall controls for the activities.
I Regional management found no regulatory basis for a violation with regard to either issue.
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Region IV Management Handling of the Issue l
1 Although this item could have been dropped from the report entirely, it was carried as part of the unresolved iten discussed under item 7 above,
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CPSES ISSUE DATE PROCEDITRE NO.
NUCLEAR OPERATIONS ENGINEERING MANUAL QCJ g g N0E-205 LICENSING REVISION NO. 1 PAGE 2 0F 11
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1.0 Purpose The purpose of this procedure is to provide a documented method for 1
reviewing and responding to NRC licensing correspondence assigned to
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Nuclear Operations by the Nuclear Licensing Supervisor. This procedure l also describes the overall process for coptrollin'g and coordinating revisions to CPSEC licensing documentti v.
2.0 Applicability This procedure is applicable to licensing correspondence assigned to Nuclear Operations for review. This procedure is also applicabla to revisions to Comanche Peak Steam Electric Station (CPSES) licetsing documents. The delineation of responsibilities and organizational interfaces related to the revision of licensing documents applies to Nuclear Operation staff; however, the process described includes work functions performed by other organizations. Specific responsibilities within organizations outside the Nuclear Operations staff are defined in accordance with that organizations licensing procedures. This procedures becomes effective when issued.
3.0 Definitions 3.1 Licensing Correspondence - Those letters, transmittals and memorandums which transmit information of a technical or admini-strative nature concerning licensing and regulatory matters.
4.0 Instructions 4.1 Licensing Correspondence Review Responsibilities 4.1.1 The corporate Nuclear Licensing Supervisor will be respon-l sible for forwarding all applicable licensing correspon-dence to the Operations Support Superintendent.
4.1.0 The Operations Support Superintendent (OSS) will be responsible for coordinating the review of the licensing correspondence. The OSS will also be responsible for coordinating the preparation of a response to licensing correspondence requiring a TUCCo response and prepare the response for the appropriate signature in a timely menner.
4.1.3 The Operations Support En ineer nat nos Circulars in accordance with Procedure NOS-103, " Review and Assessment of Industry Operating Experiences".
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CPSES ISSUE DATE PROCEDURE NO.
N'JCLEAR OPERATIONS ENGINEERING MANUAL OCT 071985 NOE-205 LICENSING REVISION NO. 1 PAGE 3 0F 11 4.1.4 The Technical Support Engineer (or his designate, here-l after referred to as TSE) shall be responsible for reviev-ing all Inspection and Enforcement Bulletins and other licensing correspondence auch as, but not limited to the Federa *
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.n o,,. :i 4.2 Licensing Correspondence Review Proce6s 4.2.1 The OSS will forward all IE Bulletins and other licensing correspondence mentioned in Section 4.1.4 to the Technical Support Section.
4 4.2.2 lE Bulletins which do not require a response vill be forwarded to the Operations Support Engineer for inclusion in the 10ER Program in accordance with Procedure NOS-103.
IE Bulletins which require a response vill be handled by the Licensing Correspondence Review Process as described below.
4.2.3
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A distinct numerie identifier, the Licensing Correspondence Review Number (LCR No.), vill be assigned to the licensing correspondence. The correspondence vill be entered in the LCR Log.
4.2.4 For each licensing correspondence logged, a Licensing Correspondence Review Form (Attachment 2), will be attached.
4.2.5 The first section of the Licensing Correspondence Review Form (LCR Form) vill be completed by entering the type of licensing correspondence, the LCR No., a brief description of the subject of the correspondence, date received, and when the response is due.
4.2.6 The TSE shall assip a Technical Support Section engineer to review the licensing correspondence for requirement 4 or information within the scope of presert or future plant operational activities. After completion of the licensing correspondence review, the assigned engineer shall com-plete the LCR Form by:
1 4.2.6.1 Indicating whether or not the correspondence is applicable to CPSES by checking the appropriate box. If the correspondence is applicable, the j
engineer shall explain how it is applicable in the appropriate space.
4.2.6.2 Indicating whether any further action is re-quired by checking the appropriate box. If
l l
CPSES-ISSUE DATE PROCED71tE NO.
HUCLEAR OPERATIONS ENGINEERING MANUAL OCT 071985 NOE-205 LICENSING REVISION NO. 1 PAGE 4 0F 11 further action is required, the engineer shall state what actions must be done and develop a plan for completing those actions on the LCR Form.
4.2.7 The assigned engineer should pomplete'the review in a timely manner, sign, date end return the LCR to the TSE.
4.2.8 The TSE shall evaluate the information contained in the LCR. E en the TSE concurs with the LCR, he shall sign the LCR. Information from the LCR Form shall be entered in the LCR Log as follows:
4.2.8.1 If the response contained in the LCR indicates thee CPSES plant operations will not be affected by information contained in the licensing correspondence and no further ection is required, the LCR shall be closed out by recording the LCR completion date, and the date closed and plac-ing"N/A" in the remaining spaces in the LCR Log.
4.2.8.2 If the response contained in the LCR indicates C.
that further action is required, the LCR shall remain open. The response due date, action required and action completion date will be entered as applicable in the LCR Log.
4.2.9 A copy of all LCR's pertaining to IE Bulletins will be forwarded to the Operations Suppc,rt Engineer.
4.2.10 W en the action required is to prepare a response to the i
licensing correspondence, the assigned engineer may contact Station Personnel or TUCCo Nuclear Engineerir.g for i
assistance in preparation of the response, j
l 4.7.11 Licensing correspondence responses shall be prepared for the signature of the Operations Support Superintendent and trans:mitted to the corporate Nuclear Licensing Supervisor. l 4.2.12 Upon completion of the required action, the " Action Complete Date", the "Date Closed" End the TNO memorandum l'
number used to transmit the response (when applicable) i shall be entered in the LCR Log.
4.2.13 A copy of all responses to IE Bulletins will be forwarded
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to the Operations Support Engineer for his information.
l 4.2.14 After completing the response to an IE Bulletin, a summary of the applicable operating experience will be prapared
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9
9 CPSES ISSUE DATE PROCEDURE NO.
NUCLEAR OPERATIONS ENGINEERING MANUAL g g7 NOE-205 LICENSING REVISION NO. 1 PAGE 5 0F 11 for entry into the 10ER Data Base in accordance with NOS-103-3.
4.2.15 Upon completion of the response to an IE Bulletin and completion of any applicable action items, the IE Bulletin 4.3 Revision of Operating License Documents 4.3.1 The Operations Support Superintendent shall coordinate the processing of all revisions to the CPSES FSAR made Sy Station Personnel and Nuclear Operations staff. The CPSES Engineering Department shall coordinate the processing of all revisions to the CPSES Technical Specifications.
4.3.2 Requests for changes to the FSAR and Technical Specifi-cations by Station Personnel shall be made in accordance with Procedure STA-416 " Processing of Licensing Documents".
FSAR Change Requests that are approved by the Engineering Superintendent shall be forwarded to the OSS for process-(
ing.
4.3.1.1 All FSAR Change Requests received by the OSS shall be entered in the FSAR Change Request Log (Attachment 3).
4.3.2.2 The TSE shall assign a Technical Support Section engineer to review the FSAR Change Request for applicability and correctness. The FSAR Change Request will be transmitted to Nuclear Licensing by interoffice memorandum under signature of the Operations Support Superintendent.
4.3.2.3 When the FSAR Change Request is transmitted to j
Nuclear Licensing, the FSAR Change Request Log l
j shall be closed out by entering the date the change request was sent and the associated TNO memorandum number.
4.3.3 Request for changes to the FSAR by Nuclear Operations Staff shall be made by completing Form TND-LI-15.
4.3.3.1 The change request shall consist of a marked up copy of the original document reflecting the proposed change attached to the completed change request form.
i
CPSES ISSUE DATE
' PROCEDURE NO.
NUCLEAR OPERATIONS ENGINEERING MAm1AL OCT 0 e 166E N0E-205
~
LICENSING REVISION NO. 1 FACE 6 0F 11 4.3.3.2 The initiator shall obtain the appropriate section supervisor approval and transmit the change request to the TSE.
4.3.3.3 The change request shall'be entered into the FSAR Change Request Log. d'he TSE shall assign a Technical Suppert Section engineer to review the change request ahd forward the change request to the OSS for final approval.
4.3.3.4 The OSS shall review the change request and supporting documents. He shall signify approval on the change request form by signature and date.
- 4. 3. 3. 5 The FSAR Change Request will be transmitted to Nuclear Licensing by interoffice memorandum l
under signature of the OSS.
4.3.3.6 When the FSAR Change Request is transmitted to Nuclear licensing, the FSAR Change Request Log
(
shall be closed out in accordance with Section 4.3.2.3.
- 4. 3. 4 Request for changes to the Technical Specifications by Nuclear Operations Staff shall be made by completing Form STA-408-1.
4.3.4.1 The change request shall consist of a copy of j
the original document marked to reflect the proposed change and attached to the completed change request.
4.3.4.2 The initiator shall forward the change request to the TSE. The TSE shall review the request and recommend approval by signing and dating the change request and forwarding the change request to the OSS.
4.3.4.3 The OSS shall review the change request and supporting documents. He shall signify his final approval on the change request form by signature and date.
4.3.4.4 The OSS shall forward the Technical Specifica-tion change request to the Engineering Super-intendent.
O
__-__O
i CPSES ISSUE DATE PROCEDURE NO.
NUCLEAR OPERATIONS ENGINEERING MANUAL N0E-205 i
007 n e. ma,
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.m LICENSING REVISION NO. 1 PACE 7 0F 11 4.4 Record Retention 4.4.1 -Completed Licensing Correspondence Review forms (NOE-205 Porm B) will be retained in accordance with Procedure STA-302, " Station Records".
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4.4.2 Upon completion of the response to an IE Bulletin and complesion of any applicab1E 1tetton items, the IE Bulletin Closeout Package, including NOE-205 Fo m B, shall be retained in accordance with Procedure STA-302 " Station Records".
5.0 References 5.1 Procedure NOS-103, " Review and Assessment of Inductry Operating Experience" 5.2 Instruction NOS-103-3, " Guidance for Maintenance of Industry l
Operating Experience Report Data Base" 5.3 Procedure TND-LI-15, " Preparation, Review and Approval of FSAR Changes" 5.4 Procedure STA-416 " Processing of Licensing Documents" l
5.5 Procedure STA-501, " Reporting of Operating Information to the NRC" l
5.6 CPSES Technical Specifications, Section 6.0 " Administrative Controls" 5.7 Procedure STA-302, " Station Records" 6.0 Attachments 6.~ 1 N0E-205 Fern' A, Licensing Correspondence Review Log 6.2 N0E-205 Form B, Licensing Correspondence Review Form 6.3 NOE-205 Form C, FSAR Change Request Log 6.4 N0E-205 Form D. IE Bulletin Closecut Package L
4
CPSES ISSUE DATE PROCEDURE NO.
NUCLEAR OPERATIONS ENGINEERING MANUAL OCT 071985 N E-20$
LICENSING REVISION NO. 1 PAGE 8 0F 11, ATTACHMENT NO. 1 (PAGE 1 0F-1) g e
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CPSES ISSUE DATE PROCEDURE NO.
NUC1,EAN OPERATIONS ENGINEERING MANUAL OCT 07 M5 N E-205 LICINSING REVISION NO. 1 PAGE 9 0F 11 ATTACHMENT No. 2 (FAGE 1 of 1) tie..et c.n.e-4 ee$t.
Subjects i
note.. 4i n.e, o.ei j
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Reviewed by Date techalcel Support Eastomer 7 ste 508-205 Fors & Rev. 1 I
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CPSES ISSUE DATE PROCEDURE NO.
xvC1. EAR OPERATIONS ENGINEERING MANUAL OCT 071985
' N0E-205 LICENSING REVISION NO. 1 PAGE 10 0F 11 ATTACHMENT NO. 3 (PACE 10F 1)
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4 CP16 ISSUE AS UNDERSTOOD BY THE GEVIEW GROUP 1.
TUGCO's procedure to process Construction Deficiency Reports (CDR) failed to require file information which would give evidence of issue closure.
2.
TUSCO f ailed to revise subtier implement ing procedures before corporate NED Procedure CS-1 was issued,.resulting in conflict wi th f ive other procedures.
3.
TUGCO failed to maintain CDR files that were retrievable.
4.
TUGCO failed to report to the NRC the corrective actions actually taken and changes to commitments.
D.
TUGCG CDR files were not auditable with respect to corrective actians.
DISCUSSION:
lhe reporting requirements under 10CFR50.55(e), Construction Deficiency Reports (CDR), were instituted to provide the commission with prompt notification of significant construc ti on deficiencies. This would give the commission l
t i mel y information on which to base an evaluation of the potential safety consequences of the deficiency and det ermi ne if further regulatory action is required. [1]
I CDR's are normally identified by the licensee's quality I
assurance program through nonconformance reports, design i
deficiency reports, vendor 10CFR21 reports or other similar
- systems, i
SAFETY SIGNIFICANCE The issues that were identified by the inspector all relate to the interface between the NRC and the licensee. There is no indication that the identi fication mechani sm f or CDR's was deficient and; therefore, the sources of input to the process were f unctioning satisf actorily. This means that the deficient equipment or controlling systems were being cor r ect ed through other established mechanisms such as the nonconformance corrective action process. Thus, there is no safety significance relative to the plant equipment.
Any breakdown in the CDR reporting and tracking system would impact on the notification, evaluation and final closure as it rel ates to the NRC. The NRC requires that sel ec ted cons <t r uct i on def i c i enc y r eports be closed through i nspecti ons. [2] If detailed trackinq files are not l
mai nt aa ned, closure becomes more difficult however, the primary corrective action tracking document for the
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.i dent 1 f ed def i ci ency would be the originel quality assurance report.
CORRECTIVE ACTIONS:
- ( Dr a f t. awaiting licensee response to unresolved items) l The tesues identified in the OIA report were all either unresolved items or not transmitted to t he licensee. There would not be any formal response to the issues in the corr ecpondence to the NRC.
The licensee's construction deficiency reporting system should be structured such that they can determine when all NRC reporting requirements have been completed. Further, it l
should have a 1000 closing feature built-in to assure that i
ell-committed corrective actions have been completed as stated. This can be accomplished by reference to the appropriate formal corrective action tracking documents wi t hout me3ntalnance of duplicate fi1es.
]t wm sioted by the review group that TUGCO PROCEDURE NEO CS-1."Evaluataan of and Reporting of Items / Events Under 10EFR21 and 10CFR50.55(e)," does not specify that al1 items r epor t ed under the procedure should be first recorded in the stablished corrective action systems. The procedure states that i npiits can be received from any source. Where the source ir other than an established quali ty trac t:ing system, at is poeEible that a reported deficiency would not properly processed under a formal corrective action system.
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I" Inspection Report 50-445/85-16; 50-446,/85-13 Item 1 1.
OIA Statement of the Issue (See Attachment MM)
Issue Resolution in Final Report 1.
Failure to develop / implement Violation downgraded procedure to demonstrate 50.55(e) to unresolved deficiencies corrected.
(445/8516-U-01; 446/8513-U-01).
2.
Expanded Description of the Issue and Related Background Information
-The'01A statement of resolution is incomplete. This item was originally anLunresolved item in a previous inspection report (445/85-14; i
446/85-11). The inspector proposed to cite the licensee during this j
l inspection.
Regional management concluded that this item should remain
)
1
. unresolved.
LThe basic issue as understood by Region IV management is as follows:
1 i
i; The inspector (H. Phillips) believed that the licensee failed to l
develop'a procedure regarding the documentation to be contained in their 10 CFR 50.55(e)' files and that this failure was a violation of
-4,'
^
g(,
t t,-
V
~
2 10 CFR 50, Appendix B, Criterion V,," Instructions, Procedures, and l'
L.
Drawings." The inspector was of the view th_at Criterion V and ANSI.45.2.9 would require the licensee, in the file which contained copies of theirL50.55(e) reports to NRC, to have copies of'the records which tracked the corrective actions taken by the licensee on the issue i
in each of those reports or, alternatively, to provide a list of references to the same records.
'This' issue was initially identified in draft 1.a, paragraph 3, (Attachment 2) as a violation of Criterion V, " Instructions, Procedures, and Drawings." In draft 2a, paragraph 3a (Attachment 3), the issue is documented as unresolved. In draft 3, paragraph 4, paragraph 2f, (Attachment 5) the items remained unresolved but information concerning the TUGCo commitment to upgrade files by March 1, 1986, was addede.
It remained unresolved in the final report.
3.
Significance of the Issue I
There is no safety significance to this issue.
10 CFR 50.55(e) is a reporting requirement for significant deficiencies
)
identified during construction or design activities and does not 4
specifically address record keeping requirements. ANSI N45.2.9, Revision 11 designates only the 50.55(e) report itself as a permanent
- record, t
A
,l 1
4
~
The requirements of 10 CFR 50, Appendix B pertain ~to 10 CFR 50.55(e) in
'that the quality deficiencies that,are' identified through the various mechanisms of.the QA program should be evaluated by the appropriate licensee organization for notification and reporting as required by 10 CFR 50.55(e'). These evaluation and reporting procedures will normally
~
be an integral part of the utility.QA program. The records required to verify the compliance with 10 CFR 50 55(e), are those which demonstrate notification, evaluation, and reporting. Records necessary to track every action taken to completion resulting from a deficiency are handled q
separately within the utility's QA program.
When the final 50.55(e) report is submitted, a utility may then "close" that file, since their reporting responsibility has been satisfied.
It has been a practice of the inspection program to permit a licensee to issue a final 50.55(e) report when the corrective actions have been decided upon and the actions to be taken have been initiated via one or more of the corrective action mechanisms of the licensee's QA program.
Final action closecut is then covered by that program.
l 4.
Region IV Management Handling of the Issue l
I As discussed above,10 CFR 50.55(e) does not specify record keeping l
i requirements. Deficiencies are identified and corrected within the i
a licensee's quality assurance program and the appropriate records for these activities are maintained within that program. Therefore, Regional l
~t management' concluded that this item was not a violation.
It was lef t l
i i
il
e
. 4 4
6 Inspection Report 50-445/85-16:-50-446/85-13 f
Item 2 j
1.
DIA Statement of the Issue (See Attachment MM1 i
Issue Resolution in Final Report 2.
Failure to revise implementing Violation Downgraded procedure containing 50.55(e) to unresolved reporting.
(445/8516-U-01; 446/8513-U-01).
2.,
Expanded Description of the Issue and Related Background Information The OIA statement of resolution is incomplete. This item was originally an unresolved item in a previous inspection report 44/85-14; 446/85-11).
The inspector proposed to cite the licensee during this inspection.
Regional management concluded that this item should remain unresolved.
The basic issue as understood by Region' IV management is as follows:
The inspector (H. Phillips) believed that the failure to revise all associated implementing procedures prior to issuance of corporate procedure NE0 CS-1, " Evaluation of and Reporting of Items / Events Under i
10 CFR 21 and 10 CFR 50.55(e), was a violation of 10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings."
l 0
--__._-____._______m
OK, This issue was< initially identified in draft la,. paragraph 3, (Attachment 2) as a violation of Criterion V.
In draft 4, parag(aph 2f, (Attachment 5) this issue was changed to an unresolved item.
~
3.
Safety Significance of the Issue There-is no safety significance to this item.
Region IV management concluded that the issuance of NE0 CS-1 l
(Attachment 9) as a corporate policy prior to revising implementing l
procedures did not compromise the identification, evaluation and reporting requirements of 10 CFR 50 55(e).
It is our experience that corporate policies at all licensees are revised from time to time and that the subordinate implementing procedures are revised within some subsequent, reasonable time frame.
NE0 CS-1 was issued on October 21, 1985, and required the cognizant subordinate managers to ensure the necessary implementing procedures were created.
It is Regional management's observation that NE0 CS-1 was promptly implemented.
4.
Region IV Management Handling of the issue Regional management concluded that it was not a violation to issue a corporate policy without first revising the lower tier implementing procedures. Regional management did conclude that allowing a reasonable time frame for implementation was appro'priate.
t Il 1
8 The issue was left as unresolved pending completion of the utility commitment that revision to existing procedures would be accomplished by March 1, 1986.
It should be noted that TUGCo's corporate practice was changed s9ch that new or revised corporate policy would be issued with instructions to the subordinate managers at to the time frame allowed for revising subordinate procedures.
l l
f: '.
9 Inspection Report 50-445/85-16; 50-446/85-13' Items 3'and 5 1,
OIA Statement of the Issue (See Attachment MM) i Issue Resolution in Final Report
~
Violation'owngraded d
3.
Failure to maintain retrievable t
50.55(e) files.
to unresolved (445/8516-U-01; 446/8513-U-1).
5.
TUGCo's 50.55(e) files not Violation downgraded
~
i auditable.
to unresolved.
j l
2.
Expanded Description of the Issue and Related Background Information The basic issue as understood by Region IV management is as follows:
The inspector (H. Phillips) believed that because the "10 CFR 50.55(e) files" did not centain documentation or reference to documentation of i
final corrective action, and that after approximately 6 weeks the documentation of-final corrective action was not in the files, the
}.
utility was in violation of 10 CFR 50, Appendix B Criterion XVII, "QA Record" for failure to maintain readily retrievable QA recoEds.
i I
I.
I
__-_-_-i-___. - _ _ _. _.
u, 10 This issue was initia ly. identified in draft la, paragraph 3 (Attachment 2) as a violation of Criterion, XVII.
In draft Ib, paragraph 3 (Attachment 2), it is identified as an unresolved item.
3.
Safety Significance of the issue There is no safety significance to this issue.
i l
All 50.55(e) reports will be inspected and corrective action verified prior to licensing of CPSES as a part of the routine inspection program applicable to all reactors under construction. The closure of 50.55(e) reports will be one of the issues tracked in the Regional IE Manual Chapter 94300 letter which is updated periodically prior to licensee issuance.
Regional management also concluded that the "50.55(e) files" should demonstrate proper notification, evaluation, and reporting to demonstrate conformance to 10 CFR 50.55(e).
Records necessary to demonstrate completion of corrective action are retrievable from within the TUGCo QA records system.
This general issue was also discussed pre'viously in the discussion on item 1 for this report.
.The inspector identified the following "50.55(e) files" as containing no documentation or reference to documentation of final corrective actior.:
CP-84-27, CP-84-2', CP-85-04, CP-85-05, CP-85-11, CP-85-12, CP-85-13, and l
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(Final 50.55(e) report provided in Attachment 10.) These files were stii1 considered deficient by the, inspector after 6 weeks.
it should also be noted that Regional management has taken the position in Item 1 that copies of all corrective action records or reference to all those records in these files is not a regulatory requirement.
The licensee questioned the NRC supervisor as to whether they should concentrate on the specific 50.55(e) reports of concern to the inspector and was told that they should proceed with the overall review of 50.55(e)s they were undertaking in response to Item 1 above.
- Further, Regional management had concluded that, using the information in the existing files, it was possible to verify that corrective actions had been completed.
4.
Region IV Management Handling of the Issue i
Regional management left this issue as unresolved pending completion of the licensee's commitment to demonstrate that sufficient documentation was available to demonstrate that corrective actions were completed.
Regional management has now been informed by the utility that their review of the QA records applicable to each 50.55(e) file is essentially
- complete, it is the utility's position that QA records of completed a
correction actions were retrievable for actions described in the 50.55(e) reports. This will be verified in subsequent inspections of 50.55(e) reports.as required by MC 2512.
i
^
12 l
I Footnote
. Exception is taken by Regional management to the conclusi_on in the OIA' Summary Report which states, "..
he [the' responsible Region IV manager) did not provide adequate direction or guidance to the inspectors to obtain the additional information he believed to either develop the violation or resolve the issue...[and] (4) not directing [the inspection] to review the TUGCo QA program to identify if the QA requirement for nonconformance documents, such as 50.55(e) reports. had been satisfied.
The' utility did agree in order to fac.ilitate NRC inspection and their own audits, to provide "50.55(e) fiTes* that would provide ready l
reference to the.QA records necessary to verify corrective action.
It is managements position that these same QA records could have been located by the inspector as has been done in past inspections by 'other inspectors, but would have involved more inspection time. The direction provided to the inspector was not to expend further inspection effort, but to await the simplified files.
The comment tha.t the inspector was not directed to identify if the QA requirements for "nonconformance documents, such as 50.55(e) reports had been satisfied," is unfounded. The inspector and his consultant had been directed by Regional management to review NEO-CS-1.
In paragraph.2.f.of the final report, the review of NE0-CS-1 is documented. This same procedure was reviewed by Regional management and was, in part, the basis for the on site meeting held with TUGCo to.
discuss 50.55(e) reporting.
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13 p-Inspection Report 50-445/85-16; 50-446/85-13 Item 4 l..
1.
OIA-Statement of the Issue (See Attachment MM)
Issue Resolution in Final Report 4.
Failure to report to NRC corrective Violation downgraded actually action taken on 50.55(e)s.
to unresolved.
2.
Expanded Description of the Issue and Related Background Information The basic issue as understood by Region IV management is as follows:
The inspector (H. Phillips) interpreted the words of 10 CFR 50.55(e) that, "The report shall include... the corrective action taken.
to mean that the final 50.55(e) report should not be submitted until all corrective action has been completed. The submission of 50.55(e) reports by TUGCo stating the proposed corrective action, but before all corrective action had not been completed, was, he believed, in violation of 10 CFR 50.55(e).
In addition, four 50.55(e) reports were identified, which indicated-scheduled corrective action dates that had not been met, and three additional 50.55(e) report were identified that gave no corrective action l
L, 1- -
14
~
date. The inspector interpreted this as a failure to report on i
I corrective action, in violation;of 10 CFR 50.55(e).
This issue was initially identified in draf t la, paragraph 3 (Attachment 2) as a violation of 10 CFR 50.55(e).
In draft 2b, paragraph 4 ( Attachment 3), the issue was included as an unresolved item (in the inspector's handwriting). The issue regarding " action taken" was deleted by the inspector following discussions with Regional management as discussed below.
3.
Safety Significance of the Issue There is no. safety significance to this issue.
All 50.55(e). reports will be inspected and corrective action verified prior to licensing of CPSES as a part of the routine inspection program y
applicable to all reactors under construction. The closure of 50.55(e) reports will be one of the issues tracked in the Region IE Manual Chapter 94300 letter which is dated periodically prior to licensee issuance.
The term corrective action taken" does not imply that all physical changes have been completed. On many occasions, for example, licensees, because of long lead time procurement items, must report that the corrective action will be delayed until just prior to fuel load. There is no regulatory position regarding the interpretation taken by the t
'o i
15 inspector.
The acceptance of a final 50.55(e) report by the NRC is judgement by the inspector and Regional management.
10 CFR 50.55(e) requires "..
sufficient information to permit analysis and evaluation of the deficiency and of the corrective action.
It is not an issue in this item that sufficient information had been provided to make this determination.
The issue of corrective action date has no significance. All 50.55(e) reports will be inspected for closure.
The IE tianual Chapter 92700 (Attachment 11), states in paragraph 03.02, " Specific Guidance" that "If inspection activity identifies significant incorrect information in the report, the licensee should submit a corrected report to the NRC...
The threshold of significance of errors, includino omission, above which a corrected report is required, involves inspector judgement and should agree with considerations for citing the licensee's failure to report.
Errors of lesser significance should be discussed with the licensee for 1
accuracy of future reports." Regional management concluded that the issue of corrective action dates falls in this latter category.
I Moreover, as discussed previously, it has been the practice in the inspection program to accept a final 50.55(e) report from a licensee when the issue is being finally resolved through one of the corrective action systems of the licensee's QA program, t
.s
..t 16
'4.
Region'IV Management Handling of the Issue t
~
. Region IV management concluded'that categorizing the update of the.
scheduled da'tes'for completion of corrective actions as an unresolved 3
item was proper. The utility committed that any required updates to 50.55(e) reports would be included in the task force effort discussed i
above.
If updating does not' occur,'then some enforcement action may be l
appropriate if the items' reach the~ significance levels described above.
The issue of the phrase " action taken" was dropped based on discussion between Regional management and the inspector.
Regional management concluded this was a closed issue.
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INSPCCI!ON REPORT. 85-16/13 ISSUE i
RESOLUTION IN.FINA1. REPORT 1.Failuretodevelop/ implement p_rocedure to demonstrate 50.55(e)
Violation downgraded de7Iciencies corrected.
to unresolved (445/8516-U-01;
~
446/8513-U-01).
- 2. Failure to revise implementor procedures containing W Violation Downgraded e
~ porting.
to unresolved re (445/8516-U-01; 446/8513-U-01).
- 3. Failure to maintain retrievable 50.55(e) rTles.
Violation downgraded to unresolved (445/8516-U01; 446/8513-U-01).
- 4. Failure to report to NRC corrective actually action taken on 50.55(e)s.
Violation downgraded to unresolved.
5.TUGCO's50.55(e)filesnotaudible.
o Violation down to unresolved. graded C
- 6. TUGC0 never res'ponded to all aspects Unresolved item downgraded of IEB 79-14.
to open item 445/8516-0-03:
446/8513-0-03.
- 7. TUGCO's IEB record files were Unresolved item downgraded -
incumplete..
to open item 446/85130-053(445/8516-0-05;-
- 8. NAMC0 nwitches "EB 79-28 were not proper'y identi"ied on travelers.
Violation downgraded to Unresolved item (445/8516-U-04; 446/8513U-04).
- 9. Deficiency in TUGCO's procedures to handle IEBs.
Violation downgraded to unresolveditem(445/8516.U-02; 446/85136-U-02).
- 10. No focal point at TUGC0 to track IEB actions.
Violation / unresolved item l
dropped.
- 11. TUGC0 internal letter stated that TUGC0 not identify non-conformance Paragraph dropped in final on !!B 79-14 to NRC.
report.
- 12. 3 insufficient evidence of cuccessful nestina of BI5c0 firo see s - m ing 01 investigation requested but re.jected by Regional e
of fa' se repor1 Dy a sco - validity i
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of s uco seals cuest'oned.
management. Inspector's unresolved items maintained (445/8516-U-06.446/8513-U-06);
i 445/8516U-07:446/8513-U.07).
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%n4c, INSPECTION REPORT 85-16/13 Concern Nos. 1, 2, and 3 The'se concerns involved the applicant's system for controlling 10 CFR 50.55(e)'
documentation. As I discussed in my July 8 memorandum at pp. 23-25, it appears that violations were supportable in this area. I consider manage-ment's reasoning for downgrading the violations to be questionable.
Concern No. 4 This concern related to the improper identification of two NAMC0 switches on an installation traveler. As stated in my July 8 memoranduni at pp. 25-28, I believe that the failure to properly identify a' switch on documentation should
,be considered a violation.
Concern No. 5 This concern involved the questionable testing of BISCO electrical penetration
~
As stated in my July 8 memorandum at p. 27, I agree with :nanagement's seals.
decision to classify this item as unresolved. This item is potentially serious, however, and it must be investigated thoroughly by the staff.
In sunnary, I found seven of the 16 concerns required more information before a violation could be imposed. For nine of the concerns, however, I consider that either sufficient infonnation existed to impose a violation or l
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management's reasoning for downgrading the violation appears to be questionable.
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. item.or onLrecords traceable tc.th'e' item,.as required c
throughout. fabrication, erection, installation, and use of the: item."
Section 8.0, Identification and Control of Items (Rev. 0, 7/1/78) of the TUGCO QA Plan contains essentially.
. equivalent provisions.
c.
~ Report 85-16/13 According to P'hillips, several violations and unresolved items proposed for inspection report 85-16/13 by himself, and NRC consultants J.H. McClesky and.T.H. Young were downgraded or deleted by E.H. Johnson and T.F.
.~
- Westerman.
These proposed violations and unresolved items (according,to a matrix prepared by Phillips) and my opinion as to their validity follow:
(1)
Action on 10~CPR 50.55(e) Deficiencies 1
Proposed violation:
TUGCO failed to develop /
l implement a procedure to show or reference i
objective evidence that deficiencies were corrected.
Proposed violation:
TUGCO failed to revise implementing procedures before corporate NEO j
Procedure CS-1 was implemented, resulting in conflict with five other procedures.
Proposed violation:
TUGCO failed to maintain 50.55(e) files (QA records) that were retrievable, i.e., could not produce record in almost a month.
1 Proposed violation:
TUGCO failed to report to the NRC the corrective action actually taken and changes to commitment regarding corrective action H
reported to NRC.
(Two other proposed violations listed by Phillips in the matrix appeared encompassed by those above.)
- - - - + - - - + * - -
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In'a. handwritten note on a draft of report 85-16/13,
- j E.H. Johnson stated that 'most of the apparent violations' cited are not violations.
Records for the convenience of
(
j the NRC are not required."
The final. report indicated that
-l l
deficiencies had not been corrected by the reported date nor.
were supplemented reports provided to the NRC as to the i~
-adjusted completion dates.
The final report noted that i
TUGCO had assembled a task force.to evaluate and take action on these matters.
The item was stated to be unresolved.'
The Commission regulations in'10 CFR 50.55(e)(3) state that '[t}he [50.55(e)) report shall include a description of the deficiency, an analysis of.the safety implications and the corrective action taken, and sufficient information to 5
permit analysis and evaluation of the deficiency and of the corrective action.
If sufficient information is not available for a definitive report to be submitted within 30 days, an interim report containing all available information shall be filed, together with a statement as to when a complete report will be filed.'
Further, section 15.0, Nonconforming Items (Rev. 2, 2/18/80), of the TUGCO QA Plan requires that 50.55(e) matters be documented on a nonconformance or deficiency report.
As this information is required by the regulations and addressed in the TUGCO QA Plan, it must be available for inspection. -Thus, the proposed violations are valid as a result of the applicant's i
failure to comply with Criteria V (Instructions, Procedures,
s J'
.g 25 and Drawings), VI (Document control), and XVII (QA Records) of 10 CFR Part 50, Appendix B.
In addition, failure to report corrective action is a violation of 10 CFR 50.55(e)(3).
The violations should have been included in the final report.
Further, the final report should hot have taken credit'for the TUGCO task force in deciding to consider this item unresolved.
As side light, it is not clear to what extent E.H. Johnson considered 50.55(e) records to not be required.
It may be beneficial to have headquarters QA personnel provide guidance in this area to the regions.
~
(2)
Applicant Action on IE Bulletins Proposed unresolved item:
TUGCO never responded to all aspects of IEB-14.
Proposed unresolved item:
TUGCO IEB files for 1982 and 1985 did not contain sufficient records or reference to records which ihow IEB action / corrective action complete.
Proposed violation:
TUGCO had replaced NAMCO switches per IEB 79-28 but 2 of 14 that were field inspected were not properly identified on installation traveler.
Proposed violation: TUGCO. procedures for handling l
IEB are deficient in that they do not describe how construction management / personnel handle IEB requiring action, especially hardware repair, replacement and modification.
Proposed violation:
No TUGCO construction focal point was found for tracking IEB actions.
On the replacement of the switches, Johnson said in a handwritten note on the draft report that "we are citing I
them for a potential hardware problem on NAMCO switches --
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'A find out if there is a hardware problem or not ---otherwise there is no citation."
The final report discussed the improper identification of the two NAMCO switches on the installation travelers.
Rather than a violation, however, the report considered this item unresolved until the applicant had an opportunity to determine if other documentation'could account for this problem.
Regardless of whether other documentation was p
discovered, a violation was called for as a result of improper or inconsistent documentation of an installed component.
The final report should have considered this a violation of Criterion VIII, Identification and Control of Materials, Parts, and Components, of Appendix B to Part 50.
Additional switches should have'been inspected to ensure that this.was.an isolated incident.
i As NRC personnel cannot inspect every component or structure in the plant, it is important that the paperwork be correct to ensure that those not inspected are correct.
)
If only hardware is considered important, then a very large I
percentage of the plant hardware inust be inspected to have confidence that the plant is properly built.
This would
' include inspection of work in progress where the component or structure would become inaccessible at a later stage.
With respect to the other items related to IE Bulletins, I could not confirm any improper management action.
The two proposed unresolved items were combined as 1
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The inspector initially obtained the NAMCO switch traveller j
information from the permanent plant record vault.[1] The
'j record of events never clearly discusses why out dated records were bei ng stored nor what assurances exist that the i
switch records would ever have been updated. If the switch
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records were in the permanent plant record vault, this i ndi cat os that installation had been completed and possibly pcfion of the system had been turned over to the plant.
that Once plant recerde are completed, any changes to the
'l equipment must be identified in the records. The closecut of the unresolved item should assure that 'he licensee i
understands the root cause af the traveller mismatch with the equipment. Tpey should al so assure that the records
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system is structred such that equipment records affected by in plant equipment changes are identified and tracked to complet)on. The records system should assure tnat the changes are captured, duplicate packages do not result, and permanent records, such as equipment qualification files, j
are updated.
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tC/2'E' & 4H Inspection Report 50-445/85-16; 50-446/85-13 Item 8 1.
OIA Statement of the Issue (See Attachment MM)
Issue Resolution in Final Report 8.
NAMCO switches IEB 79-28 were not Violation downgraded to properly identified on travelers.
unresolved item (445/8516-U-04; 446/8513-U-04).
2.
Expanded Description of the Issue and Related Background Information The basic issue as understood by Region IV management is as follows:
The inspector (H. Phillips) assigned his consultant (J. McCluskey) to i
inspect a sample of 14 NAMCO switches, which had been replaced in accordance with IE Bulletin 79-28, The inspector drew the initial records (travelers) of the replacement from the Permanent Plant Records Vault and determined that 2 of the 14 were not the same model and serial number as installed. Based on this initial look, he felt that TUGCo had failed to identify and control parts in the RHR system in accordance with 10 CFR 50, Appendix B, Criterion VIII, " Identification and Control of Materials, Parts and Components."
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9 This issue was initially identified in draft la, page 4-4-(Attachment 2) as a violation of Criterion VIII, " Identification and Control of Materials, 5
i Parts, and Components." In draf t 2c, Page 4-3 ( Attachment 3), the issue was changed to unresolved pending review by TUGCo to determine if there was further documentation.
It was issued in the final report, in that
- form, paragraph 4.b (2) ( Attachment 8.a).
'3.
Safety Significance of the Issue i
This issue has no safety significance.
Initially, the inspector identified that tro NAMCO position indication switches, one on valve 1-HCV-606 and one on valve 1-FCV-618, did not match with model and serial number. Y #$
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$ tained.from,:th,e?permanentWpla d n 6,n.}Regionalmanagement i
i directed the inspector to determine whether there was later documentation on this issue and that the issue would be treated as unresolved pending that determination. Prior to the issuance of the final inspection report, the inspector informed his immediate supervisor that no further documentation had been located.
The NRC supervisor contacted the TUGCo Quality Engineering Supervisor (the normal interface point for TUGCo with the NRC on quality issues,) and inquired with regard to further documentation because this person, who had no't been contacted by the inspector on this issue. Within approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, a traveler for each valve was provided. The i
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' traveler for va5ve 1-FCV-618 matched with the NAMCO switch installed.
This switch was a safety-related switch.
The traveler for. valve-
~1-HCV-606-did not match with the NAMCO switch model number installed.
This w'as determined to'be a nonsafety-related switch, hence, there was no.
violation. The issue was carried in the final report as unresolved ~ since there'was still a question as to the use of a QA traveler to install a nonsafety device.
The ut'ility'~ personnel subsequently indicated that a QA traveler was being used in this case to prevent damage.to the NAMCO switches during installation since all NAMCO switches purchased and installed in response to to IE Bulletin 79-28 wert environmentally qualified and could, at some later time, be used in a safety application.
The inspector,-sometime later (months), carne into the Region IV trailer whir.h houses the NRC staff 'and consultants following the ongoing' CPRT l
activities, and stated that somebody was supposed to be following up on the NAMCO switch issue. The supervisor of the Comanche Peak Task Group immediately directed an inspector out of the Regioli IV Group assigned on l
site to contact the TUGCo Quality Engineering Supervisor and determine I
the disposition of the nonsafety-related NAMCO switch for valve 1-HCV-606.
A final traveler for the NAMCO switch associated with 1-HCV-606 was provided and it matched with the model number of the installed NAMCO switch. A complete documentation package was provided to the inspector to close the issue.
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-4.
Region IV Management Handling of the Issue Regional management concluded that' handling this issue as an unresolved itemlinitially was proper, since the quality documentation did not match
~
with the installed switch,-albeit in a nonsafety application.
Footnote Exception is taken to the OIA conclusion, in the summary report, that Regional management did not develop sufficient information, that the inspector was not directed "to' expand the sample size of the NAMCO switches that were inspected to ascertain if additional documentation problems existed."
Prior to issuing the report, the. documentation associated with the safety-related switch had been resolved. Had the utility been unable to produce the final traveler for the nonsafety-related switch, then an expanded sample night have been in order because then a documentation problem would have existed.. Regional management is responsible for the proper utilization of inspection resources.
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6 Fitting Locations," and CP-85-12. " Auxiliary Feedwater Pressure Control,"
stated that corrective action was scheduled to be completed for Unit 1 by.
. May 1985. The open items list indicated that they we're not completed on
' November 30, 1985.
TUGCo management has assembled a task force consisting of four to five engineers or specialists to evaluate and take action on the matters described above. Thisitemisunresolved(445/8516-U-01,446/8513-U-01).
No violations or deviations were identified.
4.
Applicant Action on IE Bulletins (IEBs) a.
In response to discussions regarding the TUGCo program concerning IEBs, Circulars, and Infomation Notices, Region IV was informed that TUGCo will' perfom a review of related procedures and records to determine the adequacy of procedures and the completeness of associated records.
The initiation of this effort will follow the task review of the 10CFRPart50.55(e)programwhichispresentlyinprogress.
Thisitemisunresolved(445/8516-U-02,446/8513-U-02).
b.
The TUGCo actions on two IEBs (i.e., Nos. 79-14 and 79-28) were selected to review hardware evaluations or repair / replacements. TUGCo l
10ER Log Sheet, page 10, dated April 9, 1984, was reviewed to determine the status of the IEBs.
(1) IEB 79-14 was evaluated by TUGCo in 1983 and was statused as closed. The NRC inspector indicated that the closure of IEB 79-14 was premature since Stone & Webster is currently analyzing Unit 1 seismic analysis versus as-built drawings, which directly relates
.to this IEB. Unit 2 as-built work has also not been completed.
l TUGCo stated that the IEB 79-14 file will be reopened and a supplemental report will be submitted upon completion of the ongoing project engineering work.
1 The above itam (IEB No. 79-14) status is considered an open item (445/8516-0-03,446/8513-0-03).
of NAMCO EA 180,11mit switches.
faulty lots, when exposed to temperatures aporized and emitted a yellow-brown crysta t can cause these switches to fail. The NRC 1 the TUSCo documentation to be complete with re specified corrective action of i switches manufactured in acceptable replacing these
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I switches on residual heat removal valves 1-HCV-606 and 1-FCV-618 c
were identified on travelers EE 82-1415-5801 and EE 83-0373-5801 as EA 180-32302 and EA 170-31302, respectively.. The switches actually installed in the field were identified as EA 180-31302 and EA 180-31302, respectively. TUGCo is evaluating this inconsistency to determine if there is other documentation to account for this.
This item is unresolved (445/8516-U-04, 446/8513-U-04).
5.
QA Records Retention
-The NRC inspectors'found that construction deficiency and IEB files were not stored in the QA records vaults. Because such records.have not been deposited in a central location, difficulties have been encountered in retrieval. TUGCo is assessing this record file issue.
This item is open pending the completion of their review (445/8516-0-05,
'446/8513-0-05).
1 6.
Electrical Penetrations In NRC Inspection Report 50-445/84-22 dated October 11, 1984, the certification of BISCO electrical penetration seals (fire barriers) was questioned with respect to the testing of the seals. During the fo11cwup of this ites, which is discussed in paragraph 2.h above, the NRC inspector identified related but different findings.
The'NRC inspector reviewed the records to determine if the documentation for eight BISCO seals support the certification statement. The eight penetrations inspected were; AB-790-174-1022A, EC-854-150A-1018A and-10188, EC-854-151A-2003A and-2004A, EC-854-1518-2025A and-2026A, and TB-803-010A-1008A.
The following documents contained apparent conflicting information that the r
NRC inspector has identified for further followup:
I BISCO letter to TUGCo dated November 13, 1984, answered the NRC certification inquiry.and stated that the subject fire barrier seal (Test No. PCA-76, ANI No. 5-26, 24"x42", floor / wall, material 6548, 9 inches depth, LAD or SLD tray, all cables, 405 loaded) met all test requirements of TUGCo Contract No. CP-0707, Gibbs & N111 Specification 2323-MS-38F, ASTM E-119, and IEEE 634 American Nuclear. Insurers (ANI) letter to BISCO dated August 20, 1985, withdrew its acceptance of 815C0 SF-20 (1977) Silicone or Dow Corning 3-6548 RTV Silicone foam for 2-and 3-hour rating without a damming board left in place. The NRC inspectors ascertained that this withdrawal was based on BISCO not having complete documentation of the test results and the recent failure of a BISCO sample tested at an independent laboratory employed by the ANI.
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_________._.______.______________.a__________
6 e.
ISSUE #8 FROM ATTACHMENT MM TO OIA REPORT 86-10
. INS'PECTION REPORT' 50-445/85-16; 50-446/85-13 6
NAMC0 switches. IEB-79-28 were not properly identified on travelers..
~
Issue As Understood by the Comanche peak Report Review Group During a routine NRC inspection of Comanche Peak Steam Electric Station-(CPSES) conducted in November 1985, the Senior Resident Inspector identified inconsistencies between certain NAMC0 switch model numbers identified on the installation instructions (travelers) and the model numbers on the installed
~hardwarein'theplant.II) The concerns raised by the Inspector with respect to this issue include deficiencies in documentation and delay associated with filing documents in the master data base'and QA vault.
In addition, in documenting the-results of its review, OIA raised a concern related to the adequacy of the hardware.(2)
DISCUSSION On December 7,1979, the NRC issued IEB 79-28, "Possible Malfunction of NAMC0 Model EA 180 Limit Switches at Elevated Temperatures". The purpose of the Bulletin was to alert the industry to a deficiency in certain manufactured lots of NAMCO EA 180 limit switches; NAMCO CONTROLS, the switch manufacture, had determined that the switch top cover gasket would emit a resin vapor at temperatures above 175'F. This vapor could condense into deposits on the normally open contacts, possibly causing a switch malfunction.
By letter dated March 24, 1980, from R. J. Gary (TUGCO) to Karl V.
Seyfrit (NRC), the applicant for CPSES responded to IEB 79-28.
In this response,
_._.________.____._____.._._____b
2 6
the applicant stated that fourteen EA-180 NAMC0 switches required replacement of the' top cover gasket, none of the switches had been put in service or exposed to ambient temperatures of more than 175'F, and that the replacement gaskets were being ordered from NAMCO and would be installed by June 30, 1980.
' Subsequently, by letter dated July 30, 1981, from R. J. Gary to Karl V. Seyfrit, the applicant revised its earlier response to IE8 7 -28.
In its revised response, the applicant stated that due to difficulty experienced resolving environmental qualification concerns, all switches within the scope of IEB
.79-28 would be replaced prior to plant operation.
During a routine NRC inspection of CPSES conducted in November 1985, the Senior. Resident Inspector focused his attention on the applicants actions in response to IEB 79-28. During the course of the inspection, the Inspector identified inconsistencies between certain NAMCO switch model numbers identified on the installation instruction (travelers) and the model numbers
-on the installed hardware. Specifically, the NAMC0 limit switches on residual heat removal (RHR) system valves 1-HCV-606 and 1-FCV-618 were identified on travelers EE 82-1415-5801 and EE 83-0373-5801 as EA 180-32302 and EA 170-31302, respectively. The switches actually installed in the field were identified as EA 180-31302 and EA 180-31302.
3 SAFETY SIGNIFICANCE 6
A.
DOCUMENTATI.ON DEFICIENCY J. Durr input on safety significance of master data base, travelers and TUGC0 inspection documentation not matching the NAMC0 switches installed in the field.
In addition, input on safety significance of not having the travelers in the master data base or QA vault for more than a year after the
' travelers were completed.
(See pages 16 and 17 of attachment 11M).
B.
HARDWARE DEFICIENCY In assessing the safety significance the Inspectors findings the CPRRG 7
analysed the worst-case assuming that the wrong switches were installed and that this situation existed without correction or recognition. Based on this review, the CPRRG determined that the worst-case scenarios are not safety significant,
'In performing the worst-case assessment to determine the safety significance associated with the NAMC0 switch concern, the CPRRG reviewed the CPSES FSAR, j
information provided by the switch manufacturer, NAMCO CONTROLS, and TU ELECTRIC, and discussed the issue with their representatives.
As shown on Figure 6.4-6 of the CPSES FSAR (3) both valves 1-HCV-606 and 1-FCV-618 are located in the "A" train of the Unit 1 system. Valve 1-FCV-618 is on [ diaphragm operated butterfly valve utilized to control by pass flow around a~
the "A" RHR heat exchanger. Valve 1-HVC-606 2 on air diaphragm operated butterfly 4
valve utilized to control discharge flow from the "A" RHR heat exchanger. The l
function of the NAMC0 switches identified by the flRC Inspector is to provide remote valve position indication with readout in the control room.
4 B.
HARDWARE DEFICIENCY - CONTINUED b
In assessing the safety significance of an error involving the installation he CPRRG evaluated the' consequences of having switch model e
number EA 180-31302 on 1-FCV-618 in lieu of switch model number EA 170-31302.
l l
Based on its review, the CPRRG determined that the switches differ in their construction and capability to function in a hostile environment. (4)
Although both EA 180 and EA 170 switches are designed to function in a high i
radiation field, the EA 180 series switches have been designed to function in containment following a design bases accident under harsh environmental conditions with elevated temperatures and pressures.. Therefore, t!ho consequence of the worst case is that a switch designed to withstand a harsh environment associated with p/st accident in-containment conditions was installed where only the capability to perform in a mild environment was required.
In assessing the safety significance of an error involving the installation of switch model number EA 180-31302 on valve 1-HVC-606 in lieu of switch model number 180-32302, the CPRRG determined the worst-case result in a non-functional valve position indicator. Based on its review, the CPRRG determined that the switches differ in the internal return spring configuration. The EA 180-31302 switches are set up at the factory to operate in the clockwise direction. (4) The consequence of this errer, if not detected, would be a false indication of valve position in the control j
i room. Although this false indication could be misleading to the plant j
operators, the CPRRG concluded that worst-case was not safety significant.
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_________m__i
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6 B.
HARDWARE DEFICIENCY - CONTINUED-This conclusion was based on other information available to the operators includingRHRdischargetothereactorcoolantsystemcoldlegtemperatfr[
recorder (TR-612) at the control board and RHR disch'arge to the reactor coolant system cold leg flow indicator (F1-618) at the control board. (3) i
.F0LLOWUP ACTIONS AND RECOMMENDATIONS i
As discussed in the OIA Report, according' to PHILLIPS, two new travelers were provided to the NRC to demonstrate the acceptability of the as-built configuration.
In addition, WESTERMAN provided 0IA copies of a TUGC0 nonconformance report and four travelers to demonstrate the acceptability of the.as-built configurations. (2) t I
l d'
6-REFERENCES 1,
Letter to Texas Utilities Generating Company from E. H. Johnsok (NRC) dated April 4,1986.
2.
DIA Report 86-10,' Attachment MM, Technical Review of PHILLIPS' Issues Contained in Comanche Peak Inspection Reports, prepared by Stephen GOLDBERG.
3.
Comanche Peak Steam Electric Station Final Safety Analysis Report, Units 1 and 2.
Docket Numbers 50-445 and 50-446.
4.
NAMC0 Limit Switches and Quick Connectors for Nuclear Environment, Series EA 180-302/602-Rev. and EA-170-302/602 Rev., 3M/5-85.
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D. of items; and that as-built hardware was adequately marked and traceable to records. The following items were randomly selected and inspected:
a.
Pressurizer Safety Valve - This item was inspected to the commitment stated in FSAR, Table 5.2-1 which includes ASME Section III, 1971 edition through winter 1972 addenda. Valve S/N N56964-00-007, which is installed in the B position was inspected. Th,e following records were reviewed:
o QA Receiving Inspection Report No. 21211 o
Code Data Report Form NV-1 o
Valve Body CMTR The valve was in place, however, installation had not been completed; therefore, the hardware installation inspection consisted of verifying tht the item was traceable to the records.
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M An A5 ect on e t'on through summer 1974 addenda, which i
is the commitment from the FSAR, Table 5.2-1.
The item was field fabricated from bulk material and installed in the CVCS with field welds number 1 and 3 (ref. BRP-CS-2-RB-076). The following records
'were reviewed:
o
- B&R Code Data Report o
Field Weld Data Card o
NDE Reports o
QA Receiving Reports (for bulk order) o Certified Material Test Report (CMTR)
The installed spool piece was inspected for weld quality and to verify that marking and traceability requirements had been met. The item had been marked with the spool piece number (3Q1) and the B&R drawing number, however, marking of the material specification number and type, heat code, or other means of traceability could not be found.
In respect to material requiring a CIGR, (nominal pipe size greater than 3/4 inch) NA-3766 requires marking with the applicable specificatfor and grade of material and heat number or heat code. When material is divided, the identification marking is required to be transferred to all pieces. This failure to identify material marking is a violation 10 CFR 50, Appendix B, Criterion VIII (446/8505-09),
c.
Loop 3 RC Cold Leg - Requirements for this item are stated in ASME,Section III, 1974 edition through summer 1974 addenda, which is the commitment from the FSAR, Table 5.2-1.
This piping subassembly l
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I 15 Inspection Report 50-445/85-07; 50-446/85-05 Item 4 1.
OIA Statement of the Issue (See Attachment MM)
Issue Resolution in Final Report 4.
For the CVCS spool piece, failure Violation dropped, to maintain traceability of item by applicable specification and grade l
of material and heat number of heat l
code.
2.
Expanded Description of the Issue and Related Background Information The basic issue as understood by Region IV management is as follows:
The NRC inspector (D. Norman) believed that the failure to mark the CVCS spool piece 3Q1(DWG No. BRP-CS-2-RB-76) with material specification and grade, heat number or heat code was a violation of Appendix B, Criterion VIII, " Identification and Control of Materials Parts, and Components," and ASME Boiler and Pressure Vessel Code,Section III, 1973 Edition, Article NA 3766.6.
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16 In the first version of NRC Inspection Report 50-445/85-07; 50-446/85-05, i,
and its associated Notice of Violatlon (Attachment 3), the inspector proposed this issue as a Level IV violation (Item 9 in NOV and paragraph i
12.a in the report). Following review and consnent by Regional management, which included the review of code' issues by regional inspector Barnes, this violation was dropped from the second version (Attachment 4)ofthereport. It does not appear in the final inspection report (Attachment 5).
3.
Safety Significance of the Issue This issue was found to have no safety significance.
As a result of the review of the first draft of the inspection report by Mr. Barnes, it was brought to Regional management's attention that marking with specification, grade, heat number, or heat code on a piping spool piece is only required during the manufacturing of the spool piece in I
accordance with ASME,SectionIII,ArticleNA3766.6(Attachment 10).
l 1
This is not necessarily a permanent marking. In accordance with NB-4122 i
(Attachment 11) pressure-retaining parts during fabrication and installation carry identification markings which will remain distinguishable until assembled or installed. It also requires an as-built sketch or a tabulation of materials identifying each piece of material with the Certified material Test Report and coded marking. The i
firstversionoftheinspectionreport(Attachment 4)statesin paragraph 14.b that "The item had been marked with the spool piece
l
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17 e
number (3Q1) and the B&P drawing number; however, marking of the material l
specification number and type, heat code, or other means of traceability could not be found." The spool piece number provides a distinguishable identification marking and the B&R drawing number (as-built sketch) provides traceability to a tabulation of mate' rials. The inspector had not properly interpreted the cod m } }l I
4.
Region IV Management Handling of Issue 1
0 The removal of this issue as a violation from the first draft of th'e inspection report was proper based on a correct interpretation of the code.
Footnote In our review of the OIA technical evaluation (Mr. Goldberg in i
Attachment MM, pages 6, 7 and 8), it is surprising to find that the testimony (Mr. Westerman under oath pages 151-154 and Mr. Barnes pages 10-14) was questioned as to the rationale for changing this finding to an unresolved item. The first version of the inspection report provides clear evidence that the spool piece number and drawing number had always appeared on the spool piece. It appears that the OIA technical reviewer failed to review the ASME Code or ask for further information in order to resolve this issue. In section MM, the OIA reviewer states that "It is unclear whether the piping subassembly identification was a factor considered by the Region when accepting the spool piece number and the drawing number, because the final report does not mention piping subassembly traceability at all." In the paragraph above this statement.
Attachment MM quoted the final inspection report, stating "The item had been marked with spool piece number (3Q1) and the Brown and ROOT (B&R) drawing number which provided traceability to the Material Certification."
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ISSUE 3.
Fai1ure to perform audits or survei11 antes of reactor pressure vessel specifications, procedures and i nst al 1 at i on.
DISCUS 510N The applicant is required to establish a quality assurance program that will function during the construction and operat3on phasrm that complies with 10CFR50, Appendix B criter.a. The qu al i t y assurance program,although designed
(
to assure prnduct q u a'i 2 t y, is also managements mechanism +0c c hec l:i ng the production process and verifying it is funct1oning proper 1y. One method of gathering the i n f or ma t i on to make this determination is the audit.
The auria t progrem shoJ1d be planned to cover those aspects of
+he t w)iiy and production process that are key i ndi c ator t of the over all process. Audits do not examine every ci,aracteristc or all elements of the process. The audii a r ei (.r ma t 1 on u h au :l d then be revi ewed by management to dot-mine if corrective actions are warranted. [2]
ihe 4 a1) ur e t a per 4 or m an audit of the reactor pressure ven el may not be significant by itself. The issue must be vi ewed )n *Pe context of the ove.al1 audit p1an to determine af the plen i s comprehensi ve. The audit is intended to s ep.1 e t 1ie quali ty f unction and verify the program ?s wori:i nq versus the quality control acti vi ti y wt'ich is
[
di r ec tl y related to determining product acceptability.
SAFETY Si FM I F I C ANCE There 2s no direct equipment safety signif2cance resulting from not auditing the installat. ion of the reactor pressure vessel. The quality of the i nstal l ati on was monitored by quality control and documented on the traveler. However, there m a'y be a broader concern if the audit plan was
. deficient or management was not reviewing the audit results and t al:ilig appropriate corrctive action. This problem has been identifed 2n NUREG 0797, Supplement No. 11, Appendix 0, Sect i on 3.2.11.
and Appendi>
P, Section 4.7.
I RECOMMENDATIONS AND FOLLOWUP The 4 ol I crn tp c4 thie assue wi11 be addressed by the Comanche Pe ed fmtponse Team as stated in NURE G 0797, Supplement 13 Abstrac t, "The NRC staff concludes that the CPRT Program 3
Plan provides en overall structure for addressing all
i e
I
-a e,
i exist 2ng and any future i ssues whi ch may be identi fi ed f roia further evaluations..."
{
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i REFERENCES 1.
Quality Control Handbook.
J.M.
Juran, Third Edition, pp 21-10 to 21-13.
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Failure to Maintain Tolerances Stated and Failure to Report These Results 7.
on a Nonconformance Report 10 CFR 50, Appendix B, Criterion XV as implemented by the TUGC0 QA Plan,Section V, Revision 2 dated May 21, 1981, requires that measures shall be established to control materials, parts or components which do not conform to. requirements; and nonconforming items shall be reviewed and accepted, rejected, repaired; or reworked in accordance with documentad procedures.
Brown and Root Quality Assurance Manual, Section 16, dated March 27, 1985, requires that unsatisfactory conditions identified on process control documents shall be identified on an Nonconforming Report.
Contraty to the above, clearances between the reaci.or vessel support brackets and support shoes were not within the tolerance stated in Construction Operation Traveler ME-79-248-55 and the condition was not reported on a Nonconformance Report.
This is a Severity Level IV Violation. (Supplement II.E) (446-8505-07)
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10 CFR 50, Appendix B, Criterion XVIM as implemented by the TUGC0 QA Plan, Section 18.0, Revision 2, dated July 31, 1984, requires that a g omorehensive system off C planned and periodic audits be carried out to verify compliance with all f,.r aspects of the quality assurance program.
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- k Contrary to the above, there was no evidence that TUGC0 had audited V p,. (,
either Unit 2 reactor vessel installation specifications, placement f
procedures, actual hardwere placement, or as-built records.
This is a Severity Level IV Violation. (Supplement II.E) (446/8505-08).
g 9.
Failure to Properly Identify a Spool Piece 10 CFR 50, Appendix B, Criterion VIII, as implemented by the TUGC0 QA Plan, Section 8.0, Revision 0, dated July 1, 1978, requires that measures be established for the identification and control of saterials, These parts, and components, including partially fabricated assembites.
measures shall assure that identification of the item ir maintained by heat number, part number or other appropriate means.
Article NA3766.6 of ASME,Section III,1974 Edition, requires that the identification of material consist of marking the material with the applicable material specification and grade of material, heat number or heat code of the material, and any additional marking required by this section to facilitate traceability of the reports of the rer.ults of all tests and examinations performed on the material.
b v'
e The QA/QC Group found an excessive number of irregularities in the inspection travelers for the fuel pool liners.
These documentation anomalies did not appear to be falsifications, but occurred because of poor QC practices.
The QA/QC Group concludes that the documentation anomalies resulted from a poor system for control of these particular travelers, and from that of a poorly implemented QC inspection program.
(See Attachment 2, QA/QC Category 6, AQ-78.)
3.2.11 $A55c se M E lE 6ategory 7)
The QA/QC Group reviewed 13 allegations in this category.
Four allegations were substantiated (AQ-20, -126, -113, -133), three were partially substantiated (AQ-69, -121, -132), and six were not (AQ-6,
-9, -25, -112, -122, -127).
Based on reviews and interviews conducted by the QA/QC Group, the allegation and con-cerns that QC was reluctant to report deficiencies in the past could not be substantiated or refuted.
In regards to the allegation of careless workman-ship, during its as-built inspections the TRT QA/QC Group found obvious care-less workmanship that QC failed to identify.
(See Attachment 2 QA/QC Category 8, AQ-50.)
With respect to the receipt of nonconforming material at CPSES, the QA/QC Group found that the receiving inspection system used at CPSES was adequate to preclude insufficiently examined or nonconforming material from being released for installation.
The QA/QC Group could not substantiate the allegation and concern in regards to i
the qualifications of B&R QA construction managers.
B&R's QA management and engineers job classifications / positions prerequisites included specific educa-tion and experience requirements.
Based on the review of selected managers qualifications (education / training), it was noted that the education require-ments for 4 upper management positions were waived using an exclusion clause.
This permitted work experience to be wholly substituted for education require-ments.
The alteration of management position prerequisites is not a violation of NRC requirements.
Nevertheless, such practice is another example of B&R overuse of the " exception to the rule clause."
The QA/QC Group substantiated the allegation and concerns of the potential for craft personnel and QC inspectors reviewing records of their own work.
Both B&R and the Anis acknowledged that past instances occurred in which record reviewers verified / accepted inspection records that contained the results of their own QC inspections.
The ANI required such records to be independently reverified.
Since record reviewers were placed in position to review their own work, the independence of record reviewers in the past is suspect.
The allegation and coM ern that QC lacked organizational independence from construction could not be substantiated or refuted.
p The QA/QC Group also concludes that improvements need to be made in the manage-ment of TUEC's exit interview program, which appeared to lack objectivity and effectiveness.
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The changes made by the QA super-visor were appropriate.
The QA C roup notes that what is important is that the auditors were inadequately trained and did not have adequate procedures to perform their audit task correctly.
The allegation and concern that TUEC management lacked commitment to an adequate QA/QC program was substantiated; e.g. failure to perform management assessment and overview of the effectiveness of the QA program and untimely reporting of significant deficiencies as required by 10 CFR Part 50.55(e).
Although TUEC's documented ' quality program manual met the NRC's requirements, the QA/QC Group found that the implementation of the QA program in a number of areas was inef-fective, because there was a lack'of senior TUEC management commitment to, and verification of, an effectively implemented QA program.
In summary, the QA/QC Group concludes that the significance and generic impli-cations of an ineffective QA program implementation are reflected in the results of the TRT's evaluation of the QA/QC programs at CPSES, including as-built inspections of completed systems or components, which had been inspected and accepted by TUEC.
(See Attachment 2, QA/QC Category 7.)
3.2.12 As-Built Issues (Category 8)
The QA/QC Group reviewed four allegations in this category.
Two allegations were not substantiated (AQ-50, -128) and two were partially substantiated (AQ-44,
-135).
With respect to the allegation and concern that craft personnel would make things fit, and NCRs were voided by engineers writing as-built or use-as-is on them, the QA/QC Group found that modifications to vendor-certified drawings, which reflected -the as-built condition, were properly decertified by the vendor's onsite representative in accordance with site procedures.
The QA/QC Group reviewed 72 NCRs that were dispositioned use-as-is and found none that were improperly dispositioned.
(See Attachment 2, Category 8, AQ-44 and AQ-128.)
The post-construction verification program (PCVP) walkdowns were made after final inspections and prior to a plant area being turned over to the TUEC startup testing organization. Walkdowns by plant operations personnel j
were not considered to be inspections, but served to identify and correct any remaining deficiencies.
The QA/QC Group could not substantiate most of the allegations and concerns relevant to the PCVP.
During the course of its review, the QA/QC Group found certain programmatic weaknesses due to a lack of guidance with respect to the level of deficiency required to initiate an NCR, I-and with respect to trending nonconformances.
The main weakness appeared to be in how to determine whether an identified nonconformance warranted more exten-sive corrective action or warranted a broader assessment for generic concerns.
(See Attachment 2, Category 8, AQ-135.)
~
The QA/QC Group pursued seven principal concerns within one allegation (AQ-50) about the as-built inspection program used by TUEC to address the NRC's Inspec-I tion and Enforcement Bulletin (IEB) 79-14, which involved verification of input used in seismic analyses for as-built safety-related piping systems.
The QA/QC Group conducted field inspections in Unit 1 in an effort to determine whether TUEC's as-built insper'. ion program functioned in proper response to f
applicable criteria of 10 CFR Part 50, Appendix B, and the requirements of Comanche Peak SSER 11 0-17 1
,7
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t?<u-M3 elm Ato J a n d&GbYMM d om A'c,oef n -rc Inspection Report 50-445/85-14/11 hsue:" "ap re+nad h" Da"4-C x _, ?__
9 0
Texas Utilities Generating Company (TUGCO) did not document an audit of records. TUGC0 performed an audit of Chicago Bridge and Iron, Inc., the-g contractor responsible for constructing the Unit No. 2 containment liner s and. failed to document the details of an audit of the quality assurance records; although, it was listed in the audit scope.I Discussion:
TUGC0 is required by 10 CFR 50, Appendix A, Criterion I, and Appendix B to establish a quality assurance program, One element of the quality assurance
)
)
program is audits of the program to insure that adequate procedures have been j
implemented. Major functions of an audit are to provide management with t
information regarding the effectiveness of the production process and the
{
i quality assurance program, and to identify nonconforming or deficient I
aspects. Based on the audit findings, management must evaluate the condition and initiate corrective actions where warranted.
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Safety Significance:
The issue is that the licensee failed to document one element of an audit, quality assurance records. The central question then becomes, was the audit 1
P I I
4
,9 4
2-1 performed'of this area. The audit record establishes that audits were Jefod 4 WMf
.u m de
\\
performed of other areas ' If you assume the audit of 4Mszerea was performed j
4 i
but not documented, then it is an administrative problem and has little j
1 consequence.
{
Aacezd6 If you assume the audit lwas not performed, then the result could be an p
4 l
inadequate records program. This has no direct safety significance and results i
%. s sg,)wc.% Aa,coz.? ?r-o 7/o.s in the same conclusion as for iterg 1 thr_ ugh 9 and 11 through 15.
A I
l Recommendations and Followup Actions:
I hpvC,.% Aen d abka~dda"/ & " ^"'
ej.dn n. 3 of s~ye Au spaa' rr-o,/o.r.
ne-ndit proce+1s is besed un somulina selected C = = of an overall ;,regrsa 4 7
to bete' odrue-pr09Pam E f f 6 C li v e n e s s. IhC I;ilLi u dew... Cut
^<T.C OSCCt DOC' w
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References:
1.
Texas Utilities Generating Company letter to Chicago Bridge and Iron, Inc., dated May 7, 1985, QXX-2381; subject: TUGC0 QA Audit Report, QA l
Audit File: TCB-6 2.
Draft Report 50-445/85-14 and 50-446/58-11, undated, CPRRG-17.
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Document Name:
r REVIEW GROUP 2/11/87 Requestor's ID:
l SHEA Author's Name:
J. Durr:ps Document Comments:
INSPECTION REPORT 50-445/85-14/11
f3 Jm %. to Jea,e 4768d M Yt Jo N d p 0 E"
/.
/
/
IbpectionReport 50-445/85-14%.s~o -vvv>/ Jaw /
~
l Issue " "- Sc., _ J t "r!'^" c-^"n V
g Texas Utilities Generating Company (TUGCO) did not document an audit of
~
records. TUGC0 performed an audit of Chicago Bridge and Iron, Inc., the contractor responsible for constructing the Unit No. 2 containment liner, and failed to document the details of an audit of the quality assurance records; although, it was listed in the audit scope.1 Discussion:
TUGC0 is required by 10 CFR 50, Appendix A, Criterion I, and Appendix B to establish a quality assurance program. One element of the quality assurance program is audits of the program to insure that adequate procedures have been
' implemented. Major functions of an audit are to provide management with information regarding the effectiveness of the production process and the quality assurance program, and to identify nonconforming or deficient aspects. Based on the audit findings, management must evaluate the condition and initiate corrective actions where warranted.
Safety Significance L su'Wumunne e The issue is that the licensee failed to document one element of an audit, quality assurance records. The central question then becomes, was the audit
,s a-I,
[
2 performed of this area. The audit record establishes'that audits were l
performed of other areas.2 If you assume the audit of this area was performed l
1 but not documented, th'n it.is'an administrative problem and has little-consequence.
i If you assume-the audit was not performed, then the result could be an inadequate records program. This has no direct safety significance and results
-in the same conclusion as for items l'through 9 and 11 through 15.
Recommendations ' and Followup ' Actions:
The audit process is based on sampling selected elements of an overall program to determine program effectiveness. The failure to document one aspect does not warrant further action.
l l
I
References:
1.
Texas Utilities Generating Company letter to Chicago Bridge and Iron, Inc., dated May 7, 1985, QXX-2381; subject: TUGC0 QA Audit Report, QA Audit File: TCB-6 2.
Draft Report 50-445/85-14 and 50-446/58-11, undated, CPRRG-17.
. 1 Y
E I
Inspect' ion Report 50-445/85-14/11 Issue as Understood by Review Group 10.- Texas Utilities Generating Company (TUGCO) did not document an audit of records. TUGC0 performed an audit of Chicago Bridge and Iron, Inc., the contractor responsible.for constructing the Unit No. 2 containment liner, and failed to document the details of an-audit of the quality assurance records; although, it was listed in the audit scope.I Discussion:
J
'TUGC0 is required by 10 CFR 50, Appendix A, Criterion I, and Appendix B to establish a quality assurance program..0ne element of the quality assurance I
4en program is audits of the program to insure that adequate procedures n
implemented /ajorfunctionsofanauditaretoprovidemanagementwith information regarding the effectiveness of the production process and the quality assurance program, and to identify nonconforming or deficient aspects. Based on the audit findings, management must evaluate the condition and' initiate corrective actions where warranted.
)
l Safety Significance:
I
-The. issue is that the licensee failed to document one element of an audit, e
)
l quality assurance records. The central question then becomes, the audit h
d
e 2
performed of this area. The audit record establishes that audits were performed of other areas.2 If you assu,ke the audit of this area was perfonned but not documented, then it is an administrative problem and has little I
consequence.
l If you assume the audit was not performed, then the result could be an inadequate records program. This has no direct safety significance and results in the same conclusion as for items 1 through 9 and 11 through 15.
s Recommendations ande(ollowup Actions:
The audit process is based on sampling selected elements of an overall program to determine program effectiveness. The failure to document one aspect does not warrant further action.
..o
)
References:
1.
Texas Utilities Generating Cornpany lette to Chicago Bridge and Iron,
l Inc., dated May 7, 1985, QXX-2381; subject: TUGC0 QA Audit Report, QA Audit File: TCB-6 2.
Draft Report 50-445/85-14 and 50-446/58-11, undated, CPRRG-17.
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l inspection Report 50-445/85-14; 50-446/85-11 i
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.OJ.A Statement of the issue (See Attachment >ti)
Issue Resolution in Final Report t
- 10. 1UGCo did not document audit Violation dropped.
N Of Tecords.
w 2.
Spuded_ Description of the Issue and Qlated Background Infomation The basic issue as understood by Region IV management is as follows:
The NRC inspector 04. Phillips) believed that TUGCo was in violation of g
I,J6TDR 50, Appendix B, Criterion XVIII ' audits" because T!4SCo ndit
' i-report TCB-6, dated May 7,1985, (Attacherent 15) of the audit of Chicago Bridge and Iron, Houston, Texas facility,, &):umented in the " audit scope" i
that an audit of 10 CFR 50, Appendix B, Celterion XVII Quality Assurance Records and the Chicago Bridge & Iron QA Manual Issue 6 sus conducted, MT h
t PAWiin4EN5fFc 4
20 r1 The issue was initially identified in draft 1, peragraph 5.c (Attachment 1) as a violation of 10 CFR 50, Appendix B, Criterion XVIII " audits." In draft 3.a. paragraph 5.c (Attachment 4) the item was changed to an unresolved item as a result of discussion with management (in the inspector's-handwriting) pending followup at the Dallatz offices of TUGCo.
In draft 4a the issue appears to have been dropptd. Regional management had not noted that it was deleted. The note un the cover of draft 4.a from Ian Barnes to the inspector requests his review. Mr. Barnes is aware of no comments from the inspector.
3.
Safety Significance of the Issue There is no safety significance to this issue.
' Regional management has concluded that this issue deals only with document detail. An audit of Chicago Bridge end Iron records was conducted by TUGCo and a summary of findings was issued.
Regional management, in reviewing with the inspector the early drafts of the report and the TUGCo audit TCB-6, had concluded that an audit of records had occurred as indicated by the existence of an issued audit report. The TUGCo letter of May 7, 1985, to Chicago Bridge and Iron concludes an over all statement of findings as follows:
l l
l
- Je :
21 j
j
" Attached is TUGCo QA Audit Report TCB-6, which describes the results l
of our audit of Chicago Bridge and Iron, Inc., Houston, Texas facility performed on April-11, 1985."
)
1
.)
l It further states:
i "No deficiencies were identified during the audit; therefore, no
]
response is required. Audit TCB-6 is considered closed."
l l
The audit summary states in the audit scope:
l "The audit'of Chicago Bridge and Iron Inc., Houston, TX was conducted i
April 11, 1985, to verify implementation of CB&I's Quality Assurance Program to assure compliance with contract number 74-2428 for the j
Unit containment liner, and the applicable criteria of 10 CFR 50 Appendix B.
The following is a list of applicable Appendix B criteria and the QA manual utilized:
l 1.
Organization II.
Quality Assurance Program IV.
Procurement Document Control V.
Instruction, Procedure, and Drawings 3
VI.
Document Control XVI. Corrective Action XVII. Quality Assurance Records j
\\
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.~
22 XVII. Audits Chicago Bridge and Iron QA Manual Issue 6."
l The audit scope as shown above clearly states that 10 CFR 50, Appendix B, l
Criterion XVII " Quality Assurance Records" was " utilized."
I In the " audit summary" section of the audit report, a summary statement (one to three sentences) was prepared for seven of the areas whicn were audited but such a summary statement was missing for the audited area of "QA records." Regional management believes this is a moot point since there were no adverse findings in this audit and the failure to include an individual statement regarding QA records appeared to be an oversight at that time.
The inspector insisted, however, that he was going to the TUGCo Dallas i
offices to review the audit file in detail. This item was therefore was therefore carried as unresolved. In the same time frame, Mr. Wells, the former TUGCo QA manager brought a memo dated January 9,1986. TUGCo (Attachment 16), which he was going to attach to the audit file of TCB-6 to clarify the detail of the records audited. Mr. Wells indicated that the information regarding the portion of audit report related to records was left out due to an administrative oversight. The records program had, in fact, been audited.
I It should be noted that this item was changed from a violation to an unresolved item, by the inspector, in draft 3a, well in advance of the
1 23 January 9 memo referred to above.
In typing draft 4a of the report l
(December 23,1985), this item was not included. Regional management has not pursued, with the inspector, his rationale for removing the item.
It is presumed that the inspector made the change.
4.
Region IV Management Handling of the Issue The classification of this issue to an unresolved item was proper pending j
the inspector's trip to TUGCo Dallas offices to look at the audit detail.
l i
Regional management did not note this issue had subsequently been deleted from the report.
Footnote:
OIA has drawn conclusions regarding this issue, which are incorrect. The conclusion in the OIA report summary states, "An example of this inconsistency was acceptance by Westerman of TUCGo's January 9, 1986, memorandum to file to document a May 1985 TUGCo audit of the QA records control system at the CB&I records facility. TUGCo's original failure to document their audit of the CB&I QA records had been discovered during the conduct of inspection 85-14/11 in October 1985. Based on the January 9, 1986, TUGCo memorandum, Westersen deleted the proposed violation citing TUGCo's failure to document their audit form the inspection report." This is wrong in that TUGCo did not fail to document their audit. The failure was limited to failing to provide an individual sumary statement of this area as they had done for the other seven areas audited.
OIA was provided a copy of TUGCo audit (TCB-6) and subsequent memo and Phillips' Matrix and all attached drafts of the inspection report. It is evident that an audit was conducted. The inspector, in his own handwriting, downgraded the issue to an unresolved item in draft 3a of the i
review process. Regional management had not noted the issue had been i
subsequently deleted when draft 4a was typed on December 23, 1985. The inspector provided no coments on draft 4a when it was sent to him by Mr. Barnes for review. All of this occurred prior to the January 9, 1986, memo.
.]
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were transmitted to Houston, Texas, without retaining a copy and CBI procedure NRP-1 (Nuclear Records Procedure) states that corrogated cardboard packing is satisfactory for small shipments and that wood should be used for large freight shipments. The procedure further states that shipping by air freight, air express, or UPS is required for shipments containing documents difficult or hepossible to replace.
The NRC inspectors interviewed the TUGC0 civil engineering Project Contact and B & R suhontract supervisor'to detemine if the CBI records were transmitted in accordance with a TUGC0 procedure as these records belong to the otner and g/wic/
removal from site wch be centro 11ed by them. They were also asked how the records were shipped and if a method of accounting for these records was described in TUGC0 procedures and followed. No TUGC0 procedure was identified and there was no record
~
of a detailed inventory list or accounting for records transmitted.
This failure to transmit records in accordance with, a procedure and afford protection ll in transit equal to temporary storage requirements'a e additional examples of violation identified in paragraph
- aboye, yasrd1 The failure to maintain control and provide accountability of removed from storage is another example of the deviation identified in paragraph above, The NRC inspector review two recent TUGC0 audits of CBI to detemine if the QA records system was audited, M 85,
)
respectively, documented audits of CBI construction organization on site and the Quality Assurance office at Houston, Texas, The NRC inspector contacted the TUGC0 Audit Group located in Dallas, Texas, to find on. if they knew how records 1
were stored on site, transmitted from site to Houston, Texas, and detemine if the l
Houston facility was adequate and had procedures to receive, control and account l
l H
1 j
4 for these records. The NRC inspector contacted an auditor. Subsequent to contacting this auditor the QA Service Supervisor contacted the NRC inspector and stated that he had met with two other TOGC0 personnel to discuss this issue and he stated that the TUGC0 decision to sen~d records to CBI was at TUGC0's own risk. The NRC inspector was not provided answers to the questions asked.
An additional reason that the NRC inspector centacted the Audit Group, was to determine
% M eG N M Ihidit? S M
((
gimet.hTnK"*20EW'MiT4%h7e why this subject was not covered in the audit.
This failure to document audit results is a violation of 10CFE Part 50 Appendix B Criterian XVIII (445/8514-
- 446/8511-
),
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ATTACHMENT 3 -
3-l
' TEXAS UTILITIES GENERATING COMPANY
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May ' 7, 1985 QXX-2381
'Mr. Ralph Kelley.
Manager Nuclear QA Chicago Bridge & Iran, Inc.
" l' '
P.O.' Box 40066-Houston, TX 77056.
COMANCHE PEAK STEAM ELECTRIC STATION TUGC0 QA AUDIT REPORT CHICAGO BRIDGE & IRON, INC.
QA AUDIT FILE: TCB l
Dear Mr. Kelley:
Attached is TUGCo QA' Audit Report TCB-6, which. describes the results of our
. audit of the Chicago Bridge '& Iron, Inc Houston, Texas facility performed on April 11, 1985. The audit team was composed of D.W. Leigh (Team Leader) and i
D.A. Ringle,.
Attachment A contains an audit summary, including attendees of the pre-audit and post-audit meetings and persons contacted during the audit.
No deficiencies were identified during the audit; therefore, no response is required. Audit TCB-6 is considered closed.
Should you have any questions, please contact D.W. Leigh at 214/979-8871.
Very truly youns,
'. E//
I' OMM/ sam j
cc:
B.R. Clements P.
Hals tead J.T. Merritt l
R.E. Camp 1
L.M. Porplewell R.D. Gentry S.A. Cooper 4
J.C. Youngblood
.,,,,,,,,,,,,r,u,,er,or.n,u,ct,,,cco m m,
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ATTACHMENT A AUDIT SUPMARY l
TCB-6 l
l l
l l
l TUGC0 QA
9 1
A::er ante - Tre Auct: Meeting
~
CA su::: No. 706/;
Ca t e +//// /f g i
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SAme T!:le Name O
1:le D~..... Na v_k TLt (,f O hQ 92 h00 C (b )
R u l Pd,,
m c., cn o s 1
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A:::n:an:e - ?:s: Au:it Meet:n:
Cate Yl// / S$
l" Name Title Na:e Tit,6
- h. m __.h3 M G O O' Sh k 8 Nb 7e t. 00#- -<* B !
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T.:CCD CA
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TCE-6 Audit Summ?ry Audit Team:
D.W Leigh - Team Leader D.A. Ringle - Auditor-In-Training
/
Personnel Contacted:
Ralph Kelley Bob Shelton Dorothy Crow Audit Scope:
i The audit of Chicago Bridge & Iron Inc., Houston, TX was conducted April 11, 1985 to verify implementation of CB&l's Quality Assurance Program to assure compliance with contract number 74-2428 the applicable criteria of 10CFR50 Appendix B.for the Unit 2 Containment Liner, and applicable Appendix B criteria and the QA Manual utilized:The following is a list of
'I.
Organization 11.
Quality Assurance Program.
IV.
Procurement Document Control V.
Instructions, Procedures and Drawings VI.
Document Control iNI25 TQ XVill.
Audits i
Chicago Bridge & Iron QA Manual Issue 6 Audit Summary:
The audit reviewed and evaluated the Chicago Bridge & Iron Quality Assurance Program in the below listed areas:
Organization was verified by review of the QA Manual and the organization chart to assure they depict the current operating structure of the company. The audit
' team assured functional responsibilities, levels of authority and lines of communication for the management who provides direction and execution of the QA The organization proved to be in compliande with the written Q'A program.
program.
Quality Assurance Program was evaluated by the review of Technical Reports submitted by the Quality Manager to the president of the company. The audit team verified implementation of the training program by the review of Level III HDE certifications and performance history files for both assigned Level III Inspectors.
The NDE performance qualification and certification was reviewed in the areas of VT, PT, MT, UT and RT.
This area appeared to be well documented j
and meets the requirements of SNT-TC-1A 1980 Edition.
TUGC0 QA-
2
'{',
I L
TCS-6 Audit Summar.v
. Audit Summary (Cont'd.):
Instructions and Procedures were verified by the review of distribution, change control, review and approval methods.
maintained and well controlled.
Distribution, recall, aWd history was Document Control was evaluated by the review of document control procedures and verification of the system used for review and distribution.
The audit team reviewed record collection letters, Microfilm Index, acknowledgement letters and transmittal letters.
This area appeared well established and executed.
Procedure Control was assessed by the review of Procedure IPA-1 Rev. 13
" Procedure Control." An interview with the Procedure Review Coordinator was conducted to verify implementation for the review and approval cycle.
Corrective Action and Nonconformance program was verified by the review of the procedure in use for approval, disposition and closure for items identified by I
the Quality Department. The audit team reviewed six.(6) corrective action reports for approvals and closure.
no nonconformance. reports issued for centractThe audit team would like to note there were 74-2428. The Corrective Action program appeared to be well established and in compliance with the QA program.
Audits were evaluated by the review of internal and external audit reports, to assure auditor qualification, documenting and reporting results and followup action on deficiencies identified in audits.
vendor,'and five (5) internal audit reports forThe audit team reviewed ten (10) 1984/1985. The audit program was found to be adequately executed.
Based on the activities and areas reviewed, the audit team has determined that Chicago Bridge & Iron has an adequate QA Program to meet the applicable require-ments of 10CFR50 Appendix B.
Dwa (O_o A D.W. Leigh (
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Teare.eader i
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TUSCO QA -
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_______-_-__--____-___-----A
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l.t TEXAS UTILITIES GENERATING COMPANY m M Y We t T( p W I N. 4 00 N O R T H (sL14 f' > T M t: rt. L.It. MI. D A L L A,p TE: t A h T S tol
'4a rch 21, 1985 QXX-2327 Mr. Ralph Kelly Manager, Quali ty Assurance Chicago Bridge & Iron l
8900 Fairbanks N. Houston Road l
Houston, TX 77240 r
COMANCHE E AK STEAM ELECTRIC STATION TUGTi QA AUDIT NOTIFICATION t,..1CAGO BRIDGE & 1RON QA AUDIT FILE: TCB-6
Dear Mr. Kelly:
This memo will serve as notification of our intent to perfonn an audit on April 11-12, 1985.
Quality Assurance Program to assure conpliance with ContractThe audit will be c 74-2428 for the Unit 2 containment liner.
The audit team will corisist of:
D.W. Leigh - Team Leader D. A. Ringle - Audi tor-In-Training We are planning for a pre-audit discussion to begin on Thursday, Apr'il 11, 1985 at 11:00 a.m.
Should you have any questions, please contact D.W. Leigh at 214/979-8871.
(/
1 W/L)
D.L. Anderson Supervisor, QA Audits DLA: mas f
cc:
B.R. Clements P. Halstead J.T. Merritt R.E. Camp L.M. Popplewell R.D. Gentry M. Warner J.C. Youngblood S.A. Cooper
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ISSUE!M
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- 1. TUGCO's procedure to process Construction Deficiency Reports (CDR) failed to require file _ inf ormation which would give evidence of issue closure.
2.
TUGCO failed'to revise subtier implementing procedures before corporate NEO Procedure CS-1 was issued, resulting in conflict with five other procedures.
- 3. TUGCO f ailed to maintain CDR files that were retrievable.
4 TUGCO 4 ailed to report to the NRC the correct ve actions actually taken and changes to commitments.
5.
TUGCO CDR files were not auditable with respect to corrective acti ons.
DISCUSSION:
4 g,f.
fde ON The reporting requirements underg 10CFR50.55(e), Construction Deficiency Reports (CDR), were instituted to provide the NRC with prompt not i f i c at i on of significant construction deficiencies. This would give the NRC timely information on which to base an evaluation of the potential safety consequences of the deficiency and determine if further regulatory-action i s required. [1] CDR's are normally identified by the licensee's quality assurance program through nonconformance reports, design deficiency reports, vendor 10CFR21 reports or other similar systems.
SAFETY SIGNIFICANCE
- The issues that were identified by the inspector all relate to the interface between the NRC and the licensee. There is no indication that the identification mechanism f or CDR's was deficient and, therefore, the sources of input to the
- ---t process were f unctioning satisf actorily. This means that the 1 fam deficient equi pment or controlling systems were being
, negsg corrected through other established mechanisms such as the nonconformance corrective action process prescribed by ~10CFR l-50, Appendix B, Criterion XVI, Corrective Actions. Thus, there is no safety significance relative to the plant equipment.
I Any breakdown in the CDR reporting and tracking system would impact on the notification, eval u ati on and final closure as it relates to the NRC. The NRC requires that selected construction deficiency reports be closed through a
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3 s -... w. > Q r L. m y vc dyc2 % mer <d&sswd e-Adkg doc ~ M dsaa
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inspections.[2] If det ai l ed tracking files are not l
mai nt ai ned, closure becomes more difficult;however, the primary corrective action tracking document for the identifed deficiency would be the original quality assurance
---er ep or t.
The prncedure identified in item No.1 does not require certain information to be retained in the licensee's tracking file which would permit the inspector to readily determine if the item had been properly closed. This makes the file unauditable f or the inspector unless there are cross references to the corrective actions programs.
The failure to revise subtier procedures, item No.2, results
(
in nonuniformity in the processing of CDR's, but does not necessarily affett reporting to the NRC. This also affects the interface between the NRC'and the licensee, and internal processing within the licensee's organization.
The feilure to maintain CDR files that were retrievable, item No.
3.
stems from the anspectors inablility to cross reference between the CDR files and corrective action program files. This is similar to item Nos.1 and 5 in that, files that are not retrievable are also not auditable.
The f ail ure to report corrective actions actuali y taken and any changes to commitments, item No.4,directly affects the NRC/ licensee interface. Thi s impacts the NRC's ability to perform a meaningful evaluation and any decision to take further regulatory action.
CORRECTIVE ACTIONS.
The issues discussed above were all either unr esol ved items or not transmitted to the licensee in the Region IV inspection report. Accordingly, there is no formal response to the issues in the correspondence to the NRC. Because of this, information concerning planned or actual corrective action could not be evaluated by the review team.
However, the review group believes that the licensee's construction deficiency reporting system should be structured such that they can determine when all NRC reporting requirements have been completed. Further, it should have a loop closing f eature built-in to assure that all committed corrective actions have been completed as stat ed. 0 33 This can be accomplished by reference to the appropriate formal corrective action tracking documents without maint:.: d cc of duplicate fi1es.
cNAdc4 It was noted by the revi ew group that lUGCO PROCEDURE NEO I
CS-1." Evaluation of and Reporting of Items / Events Under 10CFR21 and 10CFR50.5S(e)," does not specify that all items
t iJ Ng
. reported under the procedure should be.first recorded in the established' corrective action systems. The procedure states that 2nputs can be received from any source. Where the cource is other than an established quality tracking system, i t, i s.possi bl e. that a reported deficiency-would not be properly processed under a formal corrective' action system.
1
References:
.j 1.-10 CFR50.55(e) Statement of Consideration, 37FR6459 2.
Inspection and Enforcement Manual, I n =,p ec t i on Procedure 92700, 8/13/84 l-
. 3.
Inspection and Enforcement (1anu al, Interpretations 10CFR50.55(e)- 4/1/80 Saa. ko 1. C n~a'W Jsom. 4074 e d M +1 Jo QTA 2* W 3%~'o' 9,,L fagetr $~o - WJ~/8 s~-/6 ;.ru - WG /y7-/3 f
4 ISSUE..
.m,,JiDuv c.1 hE v RW C?O'..'"
T G~and 1__
All. reporting requirements of Inspection and Enforcement Bulletin 79-14 were not met. The reporting requirement for nonconformances, paragraph 4, were not satisfied by TUGCO.
Discussion:
Quality Assurance Aspects Inspection and Enforcement Bulletin (IEB) 79-14 was issued because several operating facilities had been identified in which the as-built configuration of the piping systems did not agree with the seismic analysis design inputs. The bulletin discussed several actions requested of the licensees including the-identification and reporting of nonconforming conditions noted during inspections that would cause safety systems to be inoperable.
e
+#
It i s obvi ous by-the wordi ng c' the ulleti that O was primarily directed to plants that held an operating license at the time.the bulletin was issued. Examples of this can be found in' paragraph 2 of the bulletin," Where nonconformances are found which effett operabi1ity of any system Operability considerations are normally associated with with operating plants with Technical Specifications aerein the i
term is defined. Paragraph 2 also discusses
... systems 1
that are normally accessible..." These are defined in the bulletin as areas of the plant that can be entered during reactor operation. Lastly, the bulletin directs that,"
Facilities holding a construction permit shall inspect safety-related systems in accordance with Items 2 and 3 and report the results within 120 days." This very explicitly i
.....i _
4 4
d e
exc lutjes const ructi on f aciliti es f rom Item 4 of the bullet.in dea}jng with nonconformances.
SAFETY SIGNIFICANCE:
There is no saf ety significance to this issue based on the fact that facilities under construction were not intended to report nonconformances to the NRC to comply with IEB 79-14.
This is further supported by a discussion with a cogni: ant Inspection and Enforcement staff member who confirmed that the intent of reporting nonconformances was to assure that operating plants took appropriate corrective actions.
CORRECTIVE ACTIONS:
T hi s item was initial 1 y discussed as an unresolved item in the draft inspection report and subsequently dropped from the final inspection report. The issue was never f ormall y transmitted to the licensee and,thus, no corrective actions would have been initiated. No iurt.her actions are warranted concerning this matter.
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1m-
'.f; CP16 50-445/85216/13 ISSUE AS-UNDERSTOOD BY THE REVIEW GROUP
- 1. TUGCO's procedure to process Construction Deficiency Reports (CDR)Efailed to require file information which would give evidence of issue' closure.
2..TUGCO failed to revise subtier implementing procedures before' corporate NEO Procedure CS-1 was issued, resulting1in conflict wi th fi ve other procedures.
-k
- 3. TUGCO failed to maintain CDR files that were retrievable.
'4. TUGCO f ailed to report :to' the NRC the corrective actions.
i
~
actually taken and changes to commitments.
- 5. 'TUGCO 'CDR fil es were not auditable with respect to corrective actions.
DISCUSSION:
The reporting requirements under 10CFR50.55(e), Construction
~ Deficiency Reports (CDR), were instituted to provide the commission with prompt notification of significant construction. deficiencies. Thi,s would give=the commission t.imel y inf ormation on which to' base an evaluation of.the potential safety' consequences.of the deficiency.and determine if further regulatory action is required. [1]
CDR's are normally. identified by the licensee's: quality-assurance ' program through nonconf ormance reports, design deficiency reports,. vendor 10CFR21 reports or other similar systems.
SAFETY SIGNIFICANCE The issues that were identified by the inspector all relate to the interface between the NRC and the licensee. There is no indication that the identification mechanism-for CDR's I
was deficient and, therefore, the sources of input to the process were f unctioning satisf actorily. This means that the deficient' equipment or controlling systems were being I
corrected through other established mechanisms such as the nonconformance corrective action process prescribed by 10CFR 50, Appendix B, Criterion XVI, Corrective Actions. Thus, l
there is no safety significance relative to the plant equipment.
1 Any breakdown in the CDR reporting and tracking system would impact on the notification, evaluation and final closure as it relates to the NRC. The NRC requires that selected
{
construction deficiency reports be closed through l
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s 4,
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inspections'[2] If detail ed tracking f iles are not l
maintained, closure becomes more difficult;however, the l
primary corrective action tracking document for the j
identifed deficiency would be the original quality assurance I
report.
)
J The procedure identified in item No.1 does not require f
certain information to be retained in the licensee #s tracking file which would permit the inspector to readily determine if the item had been properly closed. This makes the file unaudi tabl e f or the inspector unless there are cross references to the corrective actions programs.
I The f ailure to revise subtier procedures, item No.2, results in nonunifarmity in the processing of CDR"s, but does not necessarily affect reporting to the NRC. This also affects the interface between the NRC and the licensee, and internal processing within the licensee's organization.
l The f ailure to maintain CDR files that were retrievable, i tem 140.
3, stems from the inspectors i nablility to cross reference between the CDR files and corrective action program files. Inis is similar to item Nos.1 and 5 in that files that are not retrievable are also not auditable.
The failure to report corrective actions actually taken and any changes to commitments, item No.4,directly affects the NRC/ licensee interface. This impacts the NRC's ability to perform a meaningful evaluation and any decision to take further regulatory action.
CORRECTIVE ACTIONS:
The issues identified in the OIA report were all either i
unresolved items or not transmitted to the licenseein the Region IV i nspection report. There would not be any formal response to the issues in the correspondence to the NRC.
The licensee's construction deficiency reporting system should be structured such that they can determine when all NRC reporting requirements have been completed. Further, it should have a loop closing feature built-in to assure that all committsJ corrective actions have been completed as stated.[3] This can be accomplished by reference to the appropriate formal corrective action tracking documents without maintainance of duplicate files.
It was noted by the review group that TUGCO PROCEDURE NEO CS-1
" Evaluation of and Reporting of Items / Events Under 10CFR21 and 10CFR50.55(e)," does not specify that all items reported under the procedure should be first recorded in the established corrective action systems. The procedure states m_m
---m
@ COO
a e
/
that inputs can be received from any source,, Where the source is other than an established quality trackino system, it is possible that a reported deficiency would not be properly processed under a f ormal corrective action system.
References:
1.
10 CFR50.55(e) Statement of Considerations, 37FR6459
?.
Inspection and Enforcement Manual, Inspection Proce: dure 92700, 8/13/84 3.
Inspection and Enforcement Manual, Int erpretati ons 10CFR50.55(e), 4/1/80 1
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d' UNITED STATES.
4 NUCLEAR REGULATORY COMMISSION R.' ;j i
8-REGION IV '
811 RYAN PLAZA DRIVE. SUITE 1000
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ARUNGTON. TEXAS 78011 May 12, 1986 MATRIX OF DRAFTS FOR REPORT 85-16/1~
H.
S.
PHILL.IPS
' Documents:
lo b l b.
Handwritten draft for Unit 2 construction inspection.
.(Submitted Ord weel Dec. 1985) 2n B 2b First revi si ons per comment L.
on ia.
3 First draft of report.
4 First draft reviewed by management and management writes in conclusion and directed changes on pages 7,8,12 and 10, Para. 6 was revised because status changed.
Sa t, Eb Second draft and Final draft.
Final incorporates minor revisions.
6a Finni report is a composite of construction, operations, and RIV technical review team f oll owup.
E>: i t Inspector in+ormed TUGCO of vi ol ati ons on December 4, 1985.
KEY
- Original Submi ssi on
- Difference
- s Mgt directed change i mpl ement ed SUN ECT / PARAGR APH DOct.IMENT S COMMENTS I
(Inspector) la Ib 2 a b b_".
4 Sa
' Fypr t Cover Face 3
se sat Violations dropped.
l j
Action on 10CFR50.55(e)
'{
Def i ci enci es Identified j
pv the Acoli cant / Pare 3.0 k
- a. TUGCO f ailed to 4
NA
$44 Violation dropped.
j
. d evel op / i mpl ement a procedure to show or reference objective evidence that defi-1 ciencies were corrected.
Violation of Criteri on V " Procedures, Instruc-tions and Drawingr."
of 10 CFR Part 50, Appendi:: B.
(Phillips,McCleskey finding; Phillips I
. wrote the viol ation).
1
l
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5/3C/Bb
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. SUBJECT / PARAGRAPH-DOUIMENTS COMMENTS Pn 2 l
( In sp ec t or ):
la
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3 4
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TUGCO failed to re-NA
- t4 Vi ol ati on. dr opped.
vise implementing 44 procedures before corporate NEO Pro-4 l
cedure CS-1 was
)
implemented, result-ing in conflict with five other proce-
~dures.
Violation of Criterion VI " Document Control"._-(Phillips finding and wrote the violation.)
(
- c.
TUGCO failed'to main-1 sit Violation dropped.
tain 50.55(e) files (OA records) that were retrieveabl e; i.e, could not pro-duce record in almost a month.
Violation of Criterion XVII "DA Records". (McCleskey finding; Phillips wrotn the violation.)
d.
TUGCO failed to re-
$44 Violation dropped, port to the'NRCathe corrective action actually taken and changes to commitment regarding corrective.
action reported to NRC. Vi ol ati on.
(McCleskey, Phillipt finding; Phillips wrote the violation.)
- e. TUGCO failed to have Mgt changed finding a procedure. This to a positive state-violation of Criterion ment about TUGCO V was changed to an action. TUGCO unresolved statement never discussed in 1.a.
(revision 4 this commitment para. 3.)
(Phillips) with inspectors before this change was made to report.
Insert #1.
C.
. - - _ - - - _ _.. - - _ _ _ _ _ - - ~ -. - - _ - - - - _ _
i 5/12/86 SUBJECT /PARAGRAF'H DOCUMENT $'
COMMENTS Pg'7 l'
(Inspector)'
la Tahb' 3
4 5a
.S
- f. r TUGCO fil es were not-4
't Mgt. dropped pare.
auditable with respect and.subst2tuted to; correcti ve f act 2 cn.
par a. where TUGCO Violation ofJ 10 CFF<
admitted vielatiori
-Part'50.55(e) and many but was taking
'TUGCO letters which action. Insert #1.
stated-records avail-i able. Changed from.
violation. to strong paragraph.
(McCleskey)
Aeolicant Action on IE Bulletins Para. 4 a.
TUGCO never responded Statement dropped.
to all aspc<ts of Unr esol ved item IED 79-14 Unr es ol ved dropped per i tem pendi ng f ur thesr-direction.
revi ew.'(McCl eskey, Phillips)
- b. TUGCO IEB f2les.for 4-48*
Statement dropped.
1982 and'1985 did Unresol ved item not contain suffi-dropped per cient records or ref -
direction.
erence to records which show IEB action /
. corrective action complete. Unresolved item.(Phillips.
McCleskey)
- c. TUGCO had replaced
- 48 Violation dropped NAMCO switches per.
to unresolved.
IEB 79-28 but 2 of 14' that were f 2 eld inspected'were.not properly identified on installation
. traveler. Vi ol ati on of Criterion VIII
" Identification /
control of Materials.
Parts,and Components" (McCleskey, Phillips finding.
Phillips wrote the viol ati on. )
l 4
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- : i.'; N Fa l j ur + to develop / a rnpl ernent proc edure to dernonst re te CDH del 2cientaes rm e r e t i t' d.
f' D : led T. o FJi.- V e l o p a p r o c edu r tL reGbrdinQ tIie docutaentStlon tc be c c:ntai rv: d i n the CDR trict. Licensee should have doc uroe nt s i n hi s C DR fi les that trecked the cloneout of each j e, sue in lieu of on)y tracking t he r.1 in the NCR files. ( f< I V toetu o to Davis)
] C'.' l. T l d '., ".
Fv11ure U ren u: 1colei.entory procedures contoining CDR ccport:na 7 ! pl.
' a '1.! t i r i L'
E6 ~150 t ht'.:
BGEOc1 Di eCl i rup l L!ine n t i nQ p r o c e dul t!.:T tx 1 o r e,
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- i :. t ! n.
- <r,o.
J. t trocedure NEO -C5-1 wns :.
1
- r. + u r ecu n t tecLuse li caused i nconsistenclec e c l
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1O l.M. ' 'l l.. u 1 rl Ietr1eV5.b]O CDbl fi3e5 The fa21ure i r.; >:: vtt E n docurnentati on i n the CDR f i les of m.n d the recult 2 ng delav 2 n ret r2 evinQ the l
n vu - t : u o
l 3nform: 4.
'" 1 violat7on
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j,,rg g c, r e ic.r t-to NRC c or rec ti ve ec t 2 ont actually tad:en r ugar d.: ng *[3R 's i
e a elat:on The 1acensee s ubrui t ted final CDR's vith proposed corrective h e :i o,9 but 1: s f o r e all ccrrective act3ons were c ornp l e t ed I F. 5 U E.i No E TUGCD C.i.h f1]e5 wet e not Budlt&ble MG.I.>e'
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' Inc ) it.sve ' d: not t rap l y that'the,CDR's were never identifled
.i i o n ! ', t%i the tr.aciing.and corrtrel proceduren syctems were
- ", h ds:1u i. i ->
Also, the nonconf c,rniance repo"tirg/I;ystem was
- 3. 't unc ti orn n;: to'the. extent tha'. CDR'c were being'identifed
.k-l
< and " e p o r.t e d. Therefore, the concern in focused'on the litenacc/NRC. interface and not on eciuipment 4
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[lThe NF' r e qui red, that CDR'E be repor ted so that' independent
, assentuents and evaluati ons of ' the ' defi c i enc y ; c an be roade
ipd any gewfi c i rupli cati ons : be determi ned f os further NRC attions Any breakdown in the licensee's reporting or t i d t t.if o g s yt.s t e n f a.r CDR's will affect,this process. There is
,4 no ' nd f e t.v c a.gni f i c ance relative to the. plant ec;uipment at naum.im' Peal.
T!ie lacensee'n'ronconformance reporting end j
t r ? : L '.,
T V E t eh, wDu)d assure that th y ' appropr.iate corret ti ve
'W. '
Pr 9 r; t: o nG wou1d be effected, s
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"Several final 10 CFR Part 50.55(e) reports contained corrective action completion dates; however, work was neither completed by the reported date nor was a supplemental. report made to the NRC to report on significant corrective action date changes. Deficiency report for CP-84-31, ' Control Room Separation Wall,' stated that corrective action was scheduled for completion by December 15, 1984; however, it was not completed as of November 30, 1985. Similarly, deficiency reports for CP-85-11. ' Instrument Fitting Locations,' and CP-85-12, ' Auxiliary l
Feedwater Pressure Control,' stated that corrective action was scheduled to be completed for Unit 1 by May 1985. The open items list indicated that they were not completed on November 30, 1985."
JOHNSON and WESTERMAN told OIA that both of them realized there was a problem
)
with the utility's handling and control of 50.55(e)'s, that both of them met on site with the utility to discuss the problem, and that the utility j
committed to establish a task force to put the 50.55(e) records in order; this l
latter point was referred to in the final inspection report.
WESTERMAN told i
f OIA, however, that there are no requirements in the regulations for identifying'the time frame to retrieve the 50.55(e) documents, there was no urgency for these records since a decision on an operating license was not
/
pending, and the procedures that implement 50.55(e) reporting requirements do not fall under Appendix B JOHNSON admitted that TUGC0's record system was somewhat archaic'and confusing but that the utility could retrieve a record, if the right person was contacted.
Technical Evaluation of These Concerns
- In the process of performing a technical evaluation, the following plant i
documents were consulted:
1.
TUGC0 Procedure NE0 CS-1, Rev. O, " Evaluation of and Reporting of i
Items / Events Under 10 CFR 21 and 10 CFR 50.55(e), 11/1/85.
2.
TUGC0 Procedure, CP-QP-16.1, Rev. 6, "Significant Construction Deficiencies," 1/16/85.
~
3.
TUGC0 Procedure CP-QP-15.6, Rev. 3 "SDAR Status Tracking," 1/16/85.
4.
TUGC0 Nuclear Engineering Procedure TNE-AD-5, Rev. 5, " Identification of Design Deficiencies and Errors," 9/3/85.
1 1
5.
Letters Clement, TUGCO, to Hunter, NRC, 50.55(3) Reports: 4/15/85, l
CP-85-13 SDAR-173, CP-85-12, (SDAR-73), and CP-55-12 (SDAR-173); and l
4/10/85,CP-85-11(SDAR-173).
i i
Concerning the first two PHILLIPS' concerns, it appears that NEO CS-1 is f
TUGC0's administrative procedure for the evaluation of and reporting of items under 10 CFR 50.55(e). The other three procedures implement the administra-tive procedure at Comanche Peak. All of these procedures are QA-controlled procedures. During a review of these procedures, it was determined there are some inconsistencies in areas such as the responsibilities of plant and 1
corporate staff to evaluate and determine the deportability of construction i
deficiencies and to serve as the point of contact with the NRC.
It is i
understood by the utility that those deficiencies exist, and TUGC0 is e
/
16 I
currer.tly implementing corrective action. However, on the basis of the version of the procedures that were available to the inspectors at the time of the inspection, it is likely these procedures would not have satisfied the QA requirements in the document control section of TUGC0's QA program, which implements Appendix B, In the absence of a review of TUGC0's QA program, this could not be determined.
1 Regarding the third PHILLIPS concern, TUGCO's 50.55(e) files could be con-
{
sidered, essentially, part of its nonconformance file system.
In addition, j
the ANSI standards provide guidance for the retrieval times for such documents
]
l or records. Control of these files is subject to the controls placed on all l
nonconformance documents and records. For example, TNE-AD-5 addresses control of TUGC0 design deficiency reports. However, in the absence of a complete review of the control of TUGC0's nonconformance documents and. records, it could not be determined whether TUGC0 satisfied the nonconformance section or the records section of its QA program, which implements Appendix B.
Conclusion For This Concern Based on the available infomation, it is inconclusive whether a violation is appropriate for those issues because:
I 1.
the inspectors did not establish whether the requirements in the QA program for document control, records control, nonconformance control had not been satisfied; and 2.
the inspectors did not provide a regulatory basis for determining that lengthy retrieval times of 50.55(a) files are a noncompliance.
Although PHILLIPS' issues were identified in the fin 61 inspection report as unresolved items, OIA was not told, and it was not reported, that any of the above actions were or are now being addressed by the Region. Because of this, Regional management's actions in identifying PHILLIPS' issues as unresolved is contrary to IE manual requirements, identified in Chapters 0400 aad 0610, for dealing with unresolved items.
Concern No. 4 PHILLIPS was concerned that Region IV management downgraded the following document control-related violation to an unresolved item:
"TUGC0 had replaced NAMC0 switches per IEB 79-28, but two of 14 that were field inspected were not properly identified on installation traveler."
l According to the draft inspection reports, the inspector found that the master data base, the travelers, and TUGC0 inspection documentation did not match the NAMCO switches installed in the field. Specifically, these were replacement NAMCO limit switches on residual heat removal valves. According to PHILLIPS' referenced documentation, two new travelers were presented a month later to NRC to show that (1) one of the NAMC0 switches was non-safety-related and, thus, not a concern of the NRC and (2) a later version of the traveler was found to have the correct identification number. The inspectars also had a l
1
PART 50
- STATEMENTS OF CONSIDERATION of 1969, that, to the fullest extent pos-as "signincant," "not indicative of,"
Certain Nuclear Reactors Exempted from sible, the pollei e. regulations and pub-
" minor,"
- magnitude " " extensive," "f re-Licensing Requirements," has also been t
lie laws of the United States shall be quently o/ curring," etc., leads to am-adopted.
interpreted and administered in accord-biguity and vaguenes.s; (21 the provision Pursuant to the Atomic Energy Act of ance with the policies set forth in that requiring that no remt' dial action shall 1954, as amended, and sections 552 and Act. Since site preparation constitutes be taken until the Commission has been 553 of title 5 of the United States Code, a key point, from the standpoint of en-notifled of the deficiency is not neces-the following amrndments to 10 CFR vironmental impact, in connection with sary, is burdensome, and represents a Parts 50 and 115 are published as a docu-the licensinc of nuclear facilities and possible unnecessary loss of time and ment subject to cod 16 cation to be effec-materials, these arnendments will fa-money to the permit holder or his con-tive 30 days af ter pubhcation in the Fro-cilitate consideration and balancing of tractors; and (3), the rule itself is un-rut Resistn.
a broader range of realistic alternatives necessary, smce existing quality assur-and provide a more signineant rnecha-ance programs require mamtenance of nism for protecting the e,1vironment records cf the denciencies and these are 37 FR 15127 during the cartier stages of a project available to the Commission.
Pub 6shed 7/28/72 for which a facility or materials license With respect to the last comments, the Ef factive 8/27/72 is being sought.
Commission beheves estrucruting of Facuity License Application The Cornmission has found that be, R'"* 8"d h'8'ing Processes cause of the provisions in il 30.11(b),
40.14 ( b), 50.10(d), and 70.14 t b), ad-See Part 2 Statements of Connderation.
vance notice of the amendments would be contrary to the public interest in that such notice might tend to defeat their an lave een ma wtut i are purpose; therefore, good cause exists for intended to resolve some cf the problems making the amendments effective with.
out the customary 30-day notice. Ae-indicated regarding clarity and ambi-37 FR 17021 cordingly, pursuant to the Atomic En-Lutty. Other signincant changes f rom the Pubhshed 8/24/72 Ef fective 8/24/72 ergy Act of 1954, as arnended. and sec-proposed rule are:
tions 552 and 553 of title 5 of the United
- a. Examples of deficiencies have been Codes and Standards for Nuclear States Code, the following amendm*nts chminated, since these made the report.
to Title 10, Chapter 1. Code of Federal ing requirements appcar more complex Powerplants Regulations, Parts 30, 40, 50, and 70 than was actually intended On June 12, 1971, the Atomic Energy are published as a document subject to
- b. The rule has been modined to per-Co'mnission published in the Frosut codificatiott to be effective upon publica-mit construction to continue subject to RzctsTra (36 F.R.11423) amendments to tion in the Protut Rtctstra (3-21-72),
the risk of subsequent disapproval by the 10 CPR Parts 50 and 115 which added Commission.
new 1150.55a and 115.43a to establish Among other requirements in the Com-minimum quality standards for the de-37 FR 6459 mission's "Quahty Assurance Criteria for sign, fabrication, erection, construction, Pubhshed 3/30'72 Nuclear Power Plants." Appendix B to testing. and inspection of certain systems Ef fective 4/29/72 10 CFR Part 50, Criterion XVI "Corree-and components of boiling and pressur.
Actton equ re t t sign eantc ized water cooled nuclear powerplants by Reporting of Deficiencies in Design
,,d e9 y e pg ted nouiru2g conformance with appropriate and Construchon of Nuclear Power-appmpnate kvels of licensee manage-P ants ment. The following amendrnent to editions of specified published industry l
On July 27, 1970, the Atomic Energy f 50 55 of 10 CFR Part 50 requires tDe codes and standards. Sections 50.55a and holder of a construction permit for a 115.43a as adopted permitted conform.
Commission pubilshed in the Frctut.
nuckar powerplant to report the more ance with the requirements of editions ElctsTER (35 F.R.12070 for pubhc com-important of these deficiencies to the of t.he spectSed Codes. Code Casu, and ment proposed amendments to 10 CF7t Commission.
Addenda which become effective after Parts 50 and jlf " Licensing of Produc.
tion and Ut ' nation Facihtles" and " Pro-It is not the intent of the Commission the date of component order, unless the cedures for Review of Certain Nuclear to require reporting of trivial matters.
Commission has published a notice in Reactors Exempted from Licensing He-Notincation is required, hnwever, of sig-the FrosuL RscisTra that compliance quirements," which would establish um-nifiennt deficiencies in dt:..gn and con-with such requirements is unacceptable form requirements for reportmg deflet-struction. The holder of a permit for for such components.
encies occurring during nuclear power-cont.truction of a nuclear powerplant is
.Ihe CommWlon has adopted amend-plant design and construction-required to notify the Commission of ments to ll 50.55& and 115.43a which, ea h deficiency found in the processes in referencing the editions of Codes All interested persons were invited to of design, manuf acture, f abrication, in*
Code Cases, an1 Addends whoss Mquirel subrnit comments or suggestions in con.
stallation, construction, testing, and in-ments must be met, include only the naction with the proposed amendments spection whiqh, were it to have remained editions of Codes, Code Cases, and Ad-within 60 days after pubhcation of the uncorrected. could have affected ad-denda through 1971 or the winter 1971 nouce of proposed rule making in the versely the safety of operations of the addenda, as appropriate. These amend-FgoruL RecrstrR. Af ter careful cons 2d.
nuclear powerplant at any time through*
ments are considered necesury for com-eration of the comments received in re.
out the expected lifetime of the plant, p11ance with sections $52 and 553 of sponse to the notice of proposed rule and which represents either (1) a sig-title 5 of the United States Code and the making and other factors involved, the nificant breakdown in any portion of the regulations of the OfBee of the Federal Commission has adopted the amend.
quality assurance program, (2) a sig-Register pertaining to incorporation by ments in the form set out below. The nificant deficiency in final designs ap-reference (1 CFR Part 20),
amendments, as adopted, reflect a num, ber of the comments received on the pro e and e for c n 3
As new or amended editto".s of appil-e cable Codes, Cods Cases, or Addenda are notice of proposed rule making.
lon of or signiacant damage to a stro issued, the Commission w!Il review them In summary, the cornments on the pro-ture, system, or component requiring cor-and amend the provisions of Il 5035a posed rule were: (1) The extensive use rective action involving extensive effort, and 115.43a es appropriate, of many adjectives and phrases, such or (4) a significant deviation from per-Since the amendments to ll 50.55a and formance spectScations requirtng cor.
115.43a are necessary to comply with, rective action involving extensive effort.
sections 552 and 553 of 11tle 5 of the A similar amendment to i 115.43 of 10 United States Code and 1 CFR Part 20.
CFR Part 115," Procedures for Review of the Comrnission has ictmd that good cause exists for omitting notice of pro-posed rule making and public procedure September 1,1982 50-SC-15
9 10 CFR 50.55(e) 5 Issue Date:
4/1/80 Guidance - 10 CFR 50.55(e), Construction Deficienc_y Reporting 1.
PURPOSE Deficiency reporting based on the requirements of Part 50.55(e) is designe to provide the NRC staff with prompt notification and timely information of deficiencies encountered dur er lants.
o le is to 2.
DISCUSSION - GENERAL The conditions of construction permits are contained in 10 CFR 50.55.
Subpart 10 CFR 50.55(e) imposes a reporting requirement on construction i
permit (Cp) holders to report each deficiency found its design and construction which if it were to have remain uncorrected could have adversely affected the safety of operations of the nuclear facility at any time throughout the expected lifetime of the plant. Reporting is limited to deficiencies which meet certain other requirements as discussed below.
j
(.
3.
RESTATEMENT OF THE REGULATION The entire subsection of 10 CFR 50.55(e) is included here for convenience.
q f
50.55(e)(1)
If the permit is for construction of a nuclear power j
plant, the holder of the permit shall notify the Connission of each deficiency found in design and construction, which, were it to have remained uncorrected, could have affected adversely the safety of operations of the nuclear power plant at any timL throughout the expected lifetime of the plant, and which represents:
(i)
A significant breakdown in any portion of the quality assurance program conducted in accordance with the requirements of Appendix B; or (ii)
A significant deficiency in final design as approved and released for construction such that the design does not j
conform to the criteria and bases stated in the safety analysis report or construction permit; or (iii) A significant deficiency in construction of or significant damage to a structure, system, or component which will require extensive evaluation, extensive redesign, or extensive repair to meet the criteria and bases stated in
/
the safety analysis report or construction permit or to otherwise establish the adequacy of the structure, system,
^;
or component to perform its intended safety function; or
T~-
~.
)
10 CFR 50.55(e) j Issue Date: 4/1/80 t
8,
(iv)
A.significant deviation from performance specifications which will require extensive evaluation, extensive redesign, or extensive repair to establish the adequacy of a structure, system, or component to meet the criteria and bases stated in.the safety analysis report or construction permit or to otherwise establish the adequacy of the structure, system, or component to perform its intended safety function.
(2)
The holder of a construction permit shall within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notify the appropriate Nuclear Regulatory Commission Inspection and Enforcement Regional Office of each reportable deficiency.
(3)
The holder of a construction permit shall also submit a written report on a reportable deficiency within thirty (30) days to the appropriate NRC Regional Office shown in Appendix D of Part 20 of this chapter.
Copies of such report shall be sent to the Director of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. The report shall include a description of the deficiency, an analysis of the safety implications and the
')
corrective action taken, and sufficient information to permit analysis and evaluation of the deficiency and of the corrective action.
If sufficient information is not available for a definitive report to be submitted within 30 days, an interim report containing all available information shall be filed.together with a statement as to when a complete report will be filed.
(4)
Remedial action may be taken both prior to and after notification of the Division of Inspection and Enforcement subject to the risk of subsequent disapproval of such action by the Commission.
4.
APPLICABILITY Subsection 10 CFR 50.55(e) applies to the CP holder and his contractors.
The CP holder is responsible for reporting each deficiency in accordance with the criteria and requirements of 10 CFR 50.55(e). The regulation applies to design and construction and encompasses all of the activities inherent in design and construction even though they may be performed by agents, contractors, subcontractors or consultants. The CP holder must establish and implement a system that assures all reportable deficiencies are identified and reported and the reporting requirement must be imposed on his agents, contractors and subcontractors.
f I
Issue Date:
4/1/80 5
[l
. -l S.
CRITERIA FOR REPORTING a.
Deficiency (1) must have been identified, i.e., found (2) related to activities conducted as authorized by a construction permit holder (design, construction or modif hation).
(3) could adversely affect the safe operation of a facility if it '
were not corrected, i.e., it is significant (4) significant deficiency relates to one or more of the.following:
(a) breakdown in QA program (b) design released for construction (c) damage to a structure, system or component
-(
(d) construction of a structure, system, or component (e) deviation from performance specifications b.
Timeliness (1) Initial report - within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (2) Written. report - within 30 days (initial or final)
(3) Supplemental written report (s) as necessary to provide all information.
c.
Reporting Organization The CP holder is responsible for implementing instructions which will provide for licensee reporting of all reportable deficiencies identified by organizations authorized by him to conduct construction phase activities.
6.
CLARIFICATION OF 50.55(e) PHRASES a.
Could adversely affect If a deficiency meets all'the criteria and it could affect adversely
(
safe operations of the facility, it is reportable.
"Could" does not imply that it would absolutely adversely affect safe operations.
It
'(~
implies a probability that safe operations may be adversely affected if the proper conditions existed.
"At any time" means that all l
service and accident conditions of operation must be considered.
j
a m
Issue Date: 4/1/80 The fact _that a deficiency is obvious and ennld not possible oo _
uncorrected and theref ore cousa not advarealy affact safa nnavatinn aves not neaate the reautrement to forma 11v rennr+ tha daficiency if it meets the criteria of 50.55(e).
~
b.
Significance To be reportable under 10 CFR 50.55(e) a deficiency must be significant.
Significant.is interpreted as having an effect or likely to have an effect on, or influence, the safe operation of-the facility in an i
adverse manner.
1 Although "significant" is not defined in 50.55(e), it is not the intent that trivia be reported. Significance _ primarily pertains to j
operational safety and not to the cost of the corrective action.
However, as indicated below, the cost to repair or redesign provides on indicator of the term " extensive." Trivial situations such as cosmetic defects are not reportable.
The test of significance includes but is not limited to safety related I
items / activities as discussed below.
(1)
It is important to note that the regulation does not specifically state that 50.55(e) applies only to safety related structures, systems and components although this T.ay be inferred from the wording.
The 50.55(e) requirement applies to any structure, system or component (SSCs) if it contains a deficiency which were it to have remained uncorrected could have affected adversely the safety of operation of the facility. This includes those SSCs that, even if not classified as safety related, could cause or contribute to the degradation of integral plant safety as a result of an adverse interaction with safety related SSCs.
Primary examples of this are undesirable conditions or failures in a nonsafety system, structure, or component which could impact or degrade safety systems or a safety function.
The inspector must use caution in applying 50.55(e) to nonsafety SSCs and must satisfy himself that the licensee has considered l
the interactions that a deficiency in a non~ safety SSC could create.
(2) If a deficiency involves inadequate management reviews, it may l
be significant.
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Issue Date:
4/1/80 4
I
\\ c.
Extensive An item is reportable if it requires extensive evaluation to determine if it is adequate to perform its intended safety function or will not impair the accomplishment of a safety function through adverse interaction.
Extensive means the expenditure of resources (time, manpower, money)
}
to a degree disproportionate with the original design, test or i
construction expenditure. The inspector should use caution - this requires judgement and experience.
For example, the lack of extensive evaluation may be used as a justification for not reporting.. But it also may indicate an inadequate evaluation due to expense involved or a failure to consider interactions and therefore should be considered suspect.
Redesign may appear to be not extensive; the inspector should verify that all interactions and interfaces have been considered and that sufficient design margin is available.
d.
Significant Breakdown in Quality Assurance A breakdown in the QA program related to any criteria in 10 CFR 50, Appendix B, may be a reportable deficiency depending upon its significance.
{
This applies to those design and construction activities affecting the safety of plant operations, including activities such as design verification, inspection, and auditing.
For example, QA program breakdown may result from an improper identification system for safety related materials. More specifically, the implementing procedures may be incomplete or otherwise inadequate, or the execution of adequate procedures may be incomplete, improper or completely ignored.
In the latter case, not following established procedures to assure that specified quality related requirements are met, for example, may constitute a breakdown in the QA program that is reportable.
Similarly, an inadannata rernid baninn tvetam that makas it imonssible_
on_ a broad crala to detaminn whether cuality reoniramants have hoon diet. is another avamnia.
in e"rh a raca avtansiva avaluation and testinn may be required to establish that applicable requirements have been met.
~
Conversely, occasional, incomplete or otherwise inadequate records that do not indicate a significant breakdown in the QA program nor j
an unsafe condition are not considered reportable.
For example, if during site construction, delivery times (from mixing to placing) of a few of many truckloads of concrete are not recorded as required, and it can be shown by other records that requirements important to
}
safety have been met, the matter would not be reportable.
These other records may be related concrete truck trip tickets, batch plant records or acceptable test results of concrete samples representing l
concrete from these trucks. The lack of complete records in this example would not lead to unsafe plant operation, nor would it con-stitute a significant breakdown in the QA Program.
7__ _ __
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L Issue Date: 4/1/80.
[
h e.-
Notification and Reporting l
(1) Notification - Reportable Deficiency
- 10. CFR 50.55(e)(2) specifies that the CP holder shall notify the appropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of each reportable deficiency.
Notification means: (a) telephone report; (b) telegraphic report; and (c) verbal report to the NRC Regional Office af ter becoming aware of a reportable deficiency, excluding holiday or weekend elapsed times. A notification to a NRC representative present at the.CP holder's facilities does not satisfy the regulation.
The threshold for notij ication (not reporting) is considered to be within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after licensee (CP holder) becomes aware of the reportable deficiency (or potentially reportable deficiency as clarified below). Aware of the deficiency means that any cognizant licensee individual has knowledge of the deficiency as a result of:
(d) observation of condition (e) a formal submittal by any organization involved in the
.)
design, construction, evaluations or inspection of the facility (f) an informal report, or allegation, by any organization or person.
(2) Notification - Potentially Reportable Deficiency All of the deportability criteria of 50.55(e) may not be satisfied when a deficiency is initially discovered.
It is not always possible for the licensee to decide promptly during an evaluation whether the identified deficiency is reportable.
However, in most cases, significance can be partially satisfied, or sound judgement will indicate potential significance.
In these cases, it should be considered that the deficiency is a potentially reportable deficiency, and the Regional Office should be notified. The CP holder should specify that it is a potentially reportable deficiency.
The following IE position has been established to alleviate the apparent conflict between prompt notification and necessary evaluation time for those cases where an extended period of time could lapse in completing a adequate evaluation of the identified deficiency:
Notification by telephone to the Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a cognizant licensee individual becomes aware of a potentially reportable deficiency is considered acceptable.
)
A potentially reportable deficiency is considered to exist 9
-._m.___..m___.
~
Issue Date: 4/1/80 o
(x 1 when: (1) an intial prompt review of available information indicates that the problem could be significant (i.e. -
partial significance is established) but, for various reasons, additional time is required to complete the evaluation; and (2) the deficiency may be considered significant, but neither a prompt review or full evaluation can be completed within 14 days due to lack of specific information.
For example, an extensive evaluation period may exist when the licensee cannot determine without testing and analysis whether the physical properties relative to the material used for a section of reactor coolant piping were met, the licensee should promptly notify the Regional Office of this matter.
If the results of the above analysis indicates that the material is not acceptable, extensive evaluation and/or rework may be required.
If this is the case, it is clearly a reportable deficiency.
Conversely, if the analysis in the above example confirms acceptability of the material, the licensee should document these results in his records and notify the Regional Office that this deficiency was determined not to be significant based on the results of f
further analysis or investigation.
Consequently, some matters
('
which require notification may not, subsequently, require a written report.
In summary, the intent is to require a prompt notification in cases where a potentially reportable deficiency has been identified but the formal evaluation required to confirm whether the item is reportable can not be completed immediately.
(3)
Interim Report The CP holder may meet the 30 day written report requirement by submitting an interim report in lieu of the complete report if sufficient information is not available for a definitive report.
The interim report should specify:
(a) the potential problem and reference the notification (b) approach to resolution of the problem (c) status of proposed resolution (d) reasons why a final report will be delayed (e) projected completion of corrective action and submittal date of the complete report.
(
1 l
l l
l
10CFR50.55(e) l Issue Date: 4/1/80 l
' (4)
Complete Report The regulation requires that the CP holder submit a written report to the appropriate Regional office within 30 days after initial notification.
If an interim report is submitted the final report shall be due on the date connitted in the interim report. The complete report shall contain:
(a) description of the deficiency (b) analysis of the safety implications. This should include an identification of interfacing systems and possible inter-actions.
(c) corrective actions taken. Corrective actions should be sufficient to correct the deficiency and prevent future identical or similar occurrences.
To prevent future occurrences the causes of the deficiency must be fully explored and identified.
(d) sufficient information to permit analysis and evaluation
~,
of the deficiency and of the corrective action.
7.
ENFORCEMENT l
If a CP holder is aware of a reportable deficiency and it can be shown by
]
objective evidence that he has not met the time reporting requirements, then he is in noncompliance with the reporting requirement of 50.55(e) and enforcement action should be taken.
j The licensee should be encouraged to discuss " deportability" with the responsible IE inspector whenever he has a question or doubt regarding this matter.
It is appropriate for the inspector to indicate his views on whether a particular matter is reportable, but the licensee should understand that the ultimate responsibility remains with the licensee, and the inspector's judgement may change during' a future inspection wherein he has an opportunity to fully review the circumstances asso-ciated with the matter.
l l
Another aspect of this Regulation related to deportability determina-I tion pertains to judgement--judgement used by the licensee in deter-mining whether a matter is reportable. The licensee has~to make a judgement based on his (or others) evaluation / analysis.
If the licensee decides, on the basis of the above, that a matter is not I
reportable, he may have satisfied the intent of this part of the l
Regulation.
However, the inspector can exercise his option and l
challenge the licensee's decision of nonreportability. A challenge
)1f may be valid if:
s the evaluation is clearly faulty by way of omission of facts engineering or othercalculations are in error
l 10 CFR 50.55(e) 1ssue Date:
4/1/80
/
l
( l l
the evaluation is not supported by adequate records the evaluation has not considered interactions past IE experience (including that of the inspector) provide a basis as precedent for deportability the licensee has established a trend or pattern of habitually evaluating deficiencies as non-reportable
}
evaluation is performed by a person (s) or organization without expertise in the subject.
The inspector has the right and the responsibility to examine the technical validity of the licensee evaluation and if an inappropriate or unsupported decision of nonreportability has been made by the licensee, enforcement action should be considered. Regional management should review and, when valid, determine the appropriate enforcement action to take.
If there is evidence that superficial evaluations are being made to procedurally satisfy or bypass NRC requirements, strong escalated enforcement action f -
should be considered.
(MC-0800 will be changed, accordingly) 8.
RELATION TO APP. B REQUIREMENTS 10 CFR 50, Appendix B, requires procedures to be established and records maintained to handle required actions relative to resolution of identified deficiencies. Procedures and records (as in (1) and (2) below) are required to assure prompt notification and adequate reporting under 50.55(e). Means to do this should be an integral part of each licensee's QA program.
(1) Implementing Procedures Although the specific requirements of 50.55(e) are few (notify, evaluate, report), implementing procedures to assure that these requirements are met should be established by the CP holder.
For example, some means (such a's procedures or instructions) are required to assure that deficiencies found in design and construction activities delegated by the licensee to others are handled properly and reported in a timely manner to the CP holder. The procedures should assure that the evaluation of the significance of the deficiency to the safety of plant operations is performed by a person (s) with adequate expertise in the subject and that adequate management review is provided.
(2) Records Q,S { 7.M ;. y } {.) s.
J
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evaluation / analysis of all deficienci on safe operations.
k or to inform the licen records the appropriate licensee management ca hether such evaluations were made or whether the
1 l.
Issue Date: 4/1/80
, '9.
RELATIONSHIP TO 10 CFR 21 REPORTING 1
Reporting of Defects and Noncompliance (10 CFR 21) imposes a reporting i
requirement on licensees and permit holders to imediately notify the Commission of defects, in basic components or the facility which could create a substantial safety hazard. There are certain situations which can result in duplicate reporting of the same defect under 50.55(e) and Part 21 requirements. Guidance that duplicate reporting is not the intent of the NRC regulations has been promulgated via NUREG-0302, Rev.1 and in correspondence supplied to the Atomic Industrial Forum.
This guidance is reproduced below:
(1)
NUREG-0302 Rev.1 Guidance Q
Must items reported as Significant Deficiencies (under 50.55(e)) or Reportable Occurrences (under 50.36) also be reported as required in 10 CFR 21?
A.
Duplicate reporting is not required.
Care should be exer-cised, however, to assure "that the Commission has been adequately informed" (521.21b) and the information specified
)
in 521.21(b)(3) is provided should the reporting party's evaluation show that a notification is required.
Q.
How do we determine when to report a " problem" under the provisions of 50.55(e) vs the provisions of Part 217 A.
550.55(e) requires initial reporting in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.of the time licensee or his agent first identifies a significant defi-ciency. A followup report is required in 30 days.
If evaluation requires substantial time to complete, interim report (s) are acceptable.
521.21(b)(1) requires reporting within two days of when the director or responsible officer obtains information reasonably indicating a failure to comply or a defect with a written repart required within five days.
In all cases, the exercise of reasonable judgement is expected in reporting potentially reportable problems to avoid the severe penalties, which could be imposed should I
the problem turn out to be reportable.
Q.
10 CFR 50.55(e), Conditions of Construction Permits, requires that the holder of a permit notify the Commission of certain designs and construction deficiencies which are also the g
subject of 10 CFR 21.
Why has not 10 CFR 50.55(e) been i
deleted?
s
-____________________.__________d
- 1 m.
t 10 CFR'50.55(e) j{
Issue Date:
4/1/80
'r-
]
L A.
550.55(e) requires reporting that would not be reported under Part 21.
For example, 1) significant damage to.a basic component following delivery to the site is report-able under 50.55(e) and not under Part 21; and 2) a signifi-cant break down in quality assurance is reportable under 50.55(e) and not under Part 21.
Q.
Is the determination of a " defect" based on the same cri--
teria as provided in Part 50.55(e)' and/or the requirements for technical specifications for operating plants?
A.
No.
In the ' case of;the. permit holder, however, a defect reportable under Part 21 would also'be reportable under 10 CFR 50.55(e).
In the. case of the licensee some items could be reportable under Pert 21 that are not reportable as LER.
Q.
For possible problems noted under 10 CFR 50.55(e) we report to the. Consnission "possible significant deficiencies." Will
/
we be allowed to report "possible defects and noncompliance"
(
under the requirements of 10 CFR Part 21?
A.
Yes, a' report may be made during the evaluation before the
. conclusion is reached that the deviation is a defect. 'A report is not required, however, until 2 days after the responsible officer or director is informed of the conclu-1 l
sion reached as a result of the evaluation.
Q.
.It appears to us that there will be more reports filed with the Commission under the requirements of 10 CFR Part 21 than
)
under 10 CFR 50.55(_e). Does the Consnission have this same belief?
A.
No. The majority of. items subject to reporting under 50.55(e) would not fit the definition in Part 21 for a " defect" involv-l ing a " substantial safety hazard." For those cases where both 50.55(e) and Part 21 reporting requirements may apply, 3
ft is expected that permit holders will renort only under j
50.55(e) as long as they include the info.aation required J
by Part 21 to adequately infonn the. Commission.
(2) Supplemental Guidance Supplied to Atomic Industrial Forum on 0/A 15 and 16 Under 21.21(b)(1) of NUREG 0302 Rev.1
(
l
(--
The regions are authorized to use the enclosed staff positions on 10 CFR Part 21. in communications with licensees. These positions were prepared in response to inquiries from AIF and supplement I
those of NUREG 0302, Rev. 1.
In particular, until pertinent reporting regulations are amended, the staff position response to
'AIF should be used in answering licensee questions on how and when 50.55(e) reporting may be used in lieu of dual reporting under both 50.55(e).and Part 21.
Issue Date:
4/1/80
'l
) When a combined 50.55(e)/Part 21 event is reported by a licensee to l
the regional office by telephone, the region should use 150.55(e)(3) l and E21.21(b)(3) information requirements for guidance to assure that the Commission is " adequately informed." Where an event is reported under 50.55(e) and it is (subsequently) established that the event is also reportable under Part 21 the licensee should be informed that it is acceptable for the licensee to provide the information required under 121.21(b)(3) via a supplement to the initial 50.55(e) report.
(From N. Moseley to Reg. Director memo of 5/8/79 forwarding 4/26/78 letter sent to AIF)
It is the staff's position that the licensee is not required to report under Part 21 an occurrence that falls within the scope of either Part 21 or 50.55(e) or Reg. Guide 1.16 if that occurrence is reported in accordance with 50.55(e) or Reg. Guide 1.16 requirements.
In such cases, it is also the staff's position that the time requirements (oral, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> under 50.55(e) and R.G.1.16) of the reporting method used would be controlling and, for the licensee, the Part 21 reporting times would not be applicable.
(Does not change prior staff position relative to information (21.21(b)(3)) requirements) i However, a director or responsible officer of a non-licensee organization upon receiving information of a reportable defect would be subject to Part 21 reporting time requirements unless he has actual knowledge the Ccanission has been adequately i nformed.
Therefore, in those cases where a non-licensee has provided the licensee, or licensees (i.e., the defect is generic in nature) with the reportable infonnation and that information is in fact reported by the licensee (s), the non-licensee is not required to duplicate the reporting.
In this instance it is also the staff's position that the non-licensee must have actual knowledge that the reporting was exe-cuted prior to expiration of applicable Part 21 reporting time requirements before he would be relieved of reporting the defect.
It should also be noted that non-licensees are not relieved of reporting until the Commission is " adequately informed." Your attention is s' specifically directed to 521.21(b)(3)(vi).
If licensee 50.55(e) report (s) do not adequately address the generic applicability, i.e., information on all such components, which the non-licensee may' be uniquely qualified to provide, the Part 21 reporting responsibility would remain with the non-licensee for providing that part of the unreported information.
l The reverse is not true because Section 50.55(e) does not have a i
provision like that included under $21.21(b) (last sentence) to
)
relieve the licensee of reporting under 50.55(e) where he had
. actual knowledge that the Comission has been adequately informed via a Part 21 report. However, the staff has stated that where
- -. ~ -
p ^*
b
Issue Date:
4/1/80
)
f.
A i the Part 21 report -includes all information required for 50.55(e) reporting it would be acceptable for the licensee's 50.55(e) report to simply reference the previously submitted Part 21 report.
(3). Additional Guidance '- Infomation Notice' 79-30
.)
RecentIEexperience(i.e.,'enforcementissuedtoS&W,B& Wand 5 Region ~ II licensees) clarifies "The staff position pennitting.
l alternate reporting via 50.55(e)'or LER of a defect was intended to avoid duplicate reporting of the same event. The use of 1
alternate reporting methods by a licensee does not relieve him-i from assuring compliance with 10 CFR Part 21.
Therefore, each i
licensee must maintain.a system which will assure compliance with all requirements of 10 CFR Part 21 and,-in particular, in
. cases where the deficiency being-reported under an alternate method is also a ' defect', to assure that all information -
required under Part 21 is forwarded to the MC via the initial or a followup written report."
10.-
10 CFR 50.55(e) EVENT FLOW DIAGRAMS
.g k
The flow diagram on the following pages illustrate the sequence of steps and considerations relative to determining whether an identified construction i
deficiency is reportable.
Figure 1 is a duplication of the guidance previously made available to licensees via NUREG-0302, Rev. 1.
Figure 2,-incorporates the IE position for assuring prompt reporting of reportable and potentially reportable deficiencies.
I I
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Issue Date: 4/1/80 FIGURE 1 10 CFR 50.55(e)
IDENTIFIED PROBLEM I
f f
ir i f BREAKDOWN DEFICIENCY DEFICIENCY DEVIATION FROM IN IN IN PERFORMANCE QA PROGRAM FIN AL DESIGN CONSTRUCTION SPECIFICATIONS I
I I
Af JU i r DEFICIENCY ADVERSELY AFFECTS SAFETY OF OPERATIONS I
AND 1 r DEFICIENCY IS SI'GNIFICANT I
AND 1 r DEFICIENCY IS IN DESIGN OR CONSTRUCTION 1 r LICENSEE ACTION REQUIRED
,)
1
7
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Issue Date:
4/1/80-15 -.
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FIGURE 2' J
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10 CFR 50.55(e) - IE POSITION DEFICIENCY, PROBLEM OR P0TENTIALLY
)
SIGNIFICANT DEFICIENCY IDENTIFIED l
I I
l BREAKDOWN DEFICIENCY DEFICIENCY DEVIATION FROM IN OR IN OR IN OR PERFORMANCE QA PROGRAM FINAL DESIGN CONSTRUCTION SPECIFICATIONS i
3-CONCLUSION OF PROMPT CONSIDERATION l
r se it
~
COULD ADVERSELY AFFECT POTENTIAL FOR SIGNIFICANCE SAFE OPERATION -'I.E.,
IS ESTABLISHED - ADDITIONAL A REPORTABLE DEFICIENCY INFORMATION IS REQUIRED MAKE NOTIFICATION OF MAKE P0TENTIAL REPORTABLE NOTIFICATION DEFICIENCY i
' SUBMIT CONTINUE EVALUATION REPORT-l 1
REPORT SUBMIT WITHDRAWAL DEFICIENCY OF NOTIFICATION i
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____._._.m_u._
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7 10 CFR $0, Appendix B, VII q
Issue Date:. #1/80
,f
'D GUIDANCE - USE OF CERTIFICATE OF CONFORMANCE LPurpose ToLidentify specific criteria which should be used by Inspection and Enforcement. personnel for the review and evaluation of licensee management control systems' pertaining to the use of certificates of conformance in lieu of original ' records.
Discussion - 10 CFR 50, Appendix B, Criterion VII requires that documentary.
l
' evidenccJhat material and equipment conform to the. procurement requirements
.shall'be available at the nuclear power plant prior to installation or use of j
such material' and equipment.
Experience has shown that in reviewing IE Circular 78-08 and 'other. Bulletins',
j
.some licensees have not effectively-implemented the requirements of. Criterion VII, and equipment 'is being received and installed that either does not conform to the procurement documents or the documented evidence that is required prior to use does not exist or is not adequate to establish that the equipment meets requirements.
One of the principal problems has been the misuse of the certificates of conformance by the licensee and the supplier. Several root causes for this misuse appear to be: (1) inadequate specification in the procurement document g
relative to acceptance of the item, (2) inadequate audits or independent j(
inspections by the licensee or his agent to assure validity of certificates of conformance, (3) inadequate evaluation of quality documentation to assure that it does establish that' equipment received meets requiremerits, (4) inadequate practice of accepting complicated engineered items solely on certificate of con.formance documentation.
ANSI N45.2.13-1976 (or 1974 draft in WASH-1309), as supplemented or modified by Regulatory Guide 1.123, provides the minimum criteria to be satisfied during the preparation of procurement documents, surveillance of suppliers and acceptance of the item or service to assure adequate control over purchased items.
The certificate of conformance system for acceptance of purchased items represents one option that a purchaser may select to assure the item meets specified l
requirements. However, as discussed under Section 10.3.3 of ANSI N45.2.13, j
the use of certificate of conformance, by itself, for acceptance-of the item t
is lim 4+ad to items of__ simple design and those which involve standard materials _
]
processes, ana tests. ~ Therefore, J_or complex engineerea items otner~
J documentation in addition to a C0C mav be required to orovide assurance that 1
.tfie item meets rann4 ramann unioue to nuclear power 31 ants.
In any case, what I
is. required for acceptance of an item should be esta)11shed (in the procurement document or appropriate procedures) and be made known~to the receiving inspection function.
Where a certificate of conformance is used solely, or in part, as the means to assure specified requirements are met, the criteria of ANSI N45.2.13 Section 10.2 should also be met.
Examples as to the types of controls deemed appropriate include: (1) the certification system, including the
(-
procedures to be followed in filling out a certificate and the administrative i
I procedure for review and approval of the certificates (including the identification of the function and the manager thereof responsible for approval of certificates), should be described in the Purchaser's/ Supplier's quality L
~
.u 10 CFR 50, Appendix B, VIf Issue Date: 4/1/80
> assurance program, (2) means should be provided to verify the validity of Supplier certificates and the effectiveness of the certification system, such as during the performance of audits of the Supplier or independent inspection or test of the items.
Such verification should be commensurate with the Suppliers' past quality performance and safety significance of the purchased item.
Commonly used ' certifications are manufacturer's certifications that a standard material (usually consumables, such as weld rod and fly ash), if tested, would exhibit the product characteristics shown on the certification document. A tvoical certification of this type is acceptable only if the using agency can_
Qemnnstrate* that the product was manufactured under a process control system which provides for product control and process records which can establish that the product was manufactured within the characteristic limits identified on the typical certification.
Acceptance review of quality documents by or for the licensee should be accomplished by suitably indoctrinated and trained personnel, that is, personnel appropriately trained in the area being reviewed.
The IE inspector should determine whether the licensee performs meaningful review of their policies relative to the use of and in verifying the validity of certifications.
1 In addition, where a purchaser upon receipt or following use of a purchased item finds a deviation from specified requirements which has not been previously identified on the srpplier documentation, the IE inspector should assure himself that the licensee has identified a need to re-audit the suppliers system of verifying that specified requirements are met before signing the C0C.
Either by the using agency approving the process control method or by performing audits to demonstrate that the required process controls are adequate.
l 1
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It UNITED STATES i
S NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
/
Washington, D.C. 20556 j
o....
INSPECTION AND ENFORCEMENT MANUAL DEPER
)
l CHAPTER 0720 j
NRC 0FFICE OF INSPECTION AND ENFORCEMENT BULLETINS AND INFORMATION NOTICES 0720-01 PURPOSE To establish responsibilities, set forth criteria, and provide guidance on the issuance and followup of IE bulletins and information notices.
0720-02 OBJECTIVE This chapter pertains to the issuance of IE bulletins and information notices.
It describes their use and provides criteria and guidance on their preparation, distribution, response, and followup.
0720-03 DEFINITIONS / DISCUSSIONS jiBib15t1$Nd%2NNe bNiiImk
- Pb5bI, 03,01 n
requests action from and/or requests a Written ' response from licensees and permit holders regarding IIItGMlDalfDiGG@pty;Yr safeguards, or environmental significance.
Licensees or permit holders may be asked to take actions over a period of time and report such actions by letter.
However, IEBs are not intended to substitute for new or revised license conditions or other continuing requirements.*
The allotted time for response to an IEB will vary depending on the significance of information to be received and the effort required to complete the requested actions.
Written responses to IEBs are usually sent to the Regional Administrator under oath or aff;rmation according to provisions of Section 182a, Atomic Energy Act of 1954, as amended.
The original copy of the cover letters and a' copy of the reports are transmitted to the U.S.
Nuclear Regulatory Commission, Document Control Desk, Washington, D.C.
20555, for reproduction and distribution.
"If a licensee or permit holder refuses to perform an action requested in r
an IEB, a requirement for the action may be imposed by an Order, following staff evaluation.
s Issue Date:
02/21/86
)
i NRC IE BULLETINS &
0720-03.01 INFORMATION NOTICES 1
A temporary instruction (TI), will be issued concurrently R with an IEB or shortly afterwards.
The TI provides guidance R to the regional offices on the scope and depth of evaluation R and followup.
R Normally, IEBs must be forwarded to the Committee to Review Generic Requirements (CRGR) to be reviewed and approved before issuance.
Steps to obtain CRGR approval have been delineated in IE Of fice Procedure 0250, " Control of Generic Requirements for Reactor Licensees."
If emergency actions are required, IEBs may be issued without review and approval by CRGR.
A preapproved Office of Management and Budget (OMB) clearance number must be referenced in each IEB.
A specific clearance number has been granted for data collection associated with 10 CFR 50 and a separate clearance number has been granted for emergency actions (Section 0720-13).
03.02 Information Notices.
An IE information notice (IN) is issued to Ticensees or permit holders to provide information that may be relevant to safety, safeguards, or environmental issues.
In IEB may be issued subsequent to an additional some cases, an evaluation performed by the staff.
ins also may be issued promptly to inform licensees and permit holders of changes in NRC procedures, of f.he implementation of rules and regulations, or nf infractions that may be pertinent to their programs.
Review and approval by CRGR, OM8 clearance, and licensee replies to ins are not required.
0720-04 RESPONSIBILITIES AND AUTHORITIES 04.01 IE Office D.irector or Designee.
a.
Approves and signs IEBs following staff review, industry comment (when applicable), and review by the Committee to Review Generic Requirements.
b.
Approves issuance of IE8s as an emergency action that is needed to protect the health and safety of the public.
In this situation, CRGR review is not necessary and the emergency OMB clearance is used.
However, the CRGR Chair-man should be notified.
A copy of the IEB together with an estimate of the reporting burden actually imposed must be sent to the Document Management Branch (DMB) within 30 days after the IEB is issued.
Issue Date:
02/21/86. _ _. _ _ _ _ _ _ _ _ _ - _ _ - _ - - _ -
o S
J' NRC.IE BULLETINS &
INFORMATION NOTICES 0720-04.02 04.02 Director, Division of Emergency Preparedness and Engineering Response (DEPER);
a.
Designated as coordinator for all IEBs and Ihs.
b.
Concurs in all IEBs and ins.
c.
Approves and signs all ins sent to reactor facilities, including those sent to several classes of licensees when
-reactor licensees are included.
j d.
Maintaint liaison with the NRC Clearance Office [ cur-R rently Document Management Branch (DMB)] on matters R relating to use of OMB clearances.
R e.
When the emergency OMB clearance is used, provides a copy R of the IEB and an estimate of the burden actually imposed R to the DMB.
R l
f.
Arranges for distrib'ution through the DMB.
R i
04.03 Director, Division of Inspection Programs (DI).
a.
Approves and signs all ins sent to only licensed non-
[
reactor facilities, b.
Concurs in all. tis. issued pertaining to IEBs.
04.04 Chief, IE Branch Having Lead Responsibility *.
a.
Serves as focal point for receipt, evaluation, and coordi-nation of all ' draft IEBs and ins concerning matters within the branch's' purview.
b.
Promptly acknowledges receipt of draft ~ IEBs and ins by establishing liaison with the initiating organization and individual.
c.
Develops IEBs and ins, as appropriate, in response to
- events, requests from other offices, and information available at the headquarters level in accordance with requirements and guidance contained herein.
"The branch having lead responsibility is determined by the subject R matter and the technical expertise in the branch.
A branch in 01 nor-R mally would be lead on IEBs and ins that will be sent to non-reactor R licensees, but may be lead on communications that will be sent to reactor R licensees.
Likewise, a branch in DEPER e.ay have the lead on issues that R
(
affect non-reactor licensees in addition to reactor licensees.
Also, a R branch in the Division of Quality Assurance, Vendor and Technical Train-R ing Center Program may have the lead on issues identified in that divi-R sion.
R Issue Date:
02/21/86
NRC IE BULLETINS &
0720-04.04d INFORMATION NOTICES d.
Makes appropriate contacts with other cognizant groups within NRC, including 'other IE groups and the Office of Nuclear Reactor Regulation (NRR) project managers for reactor events, for the purpose ~ of obtaining information and coordination relative to the pending issuance of IEBs and ins.
e.
Coordinates information exchange with regional office having jurisdiction over ' area in which an event requiring issuance of IEB or IN occurred, to ensure that all facts are ascertained.
f.
Evaluates the need for an NRC press release concerning matters of high visibility or having potential for high public interest, and, whe're appropriate, coordinates with the Office of Public Affairs on the development and issu-ance of news media releases.
g.
Contacts - vendors, manufacturers, etc. (when appropriate),
'i whose products may be involved and/or specifically identi-fied in the IEB or IN.
If a draft is transmitted for comment, a copy of the draft also will be sent to the Public Document Room (PDR) at the same time.
h.
Ensures that all draft IEBs or ins receive adequate techni-cal reviews, including assistance from other IE technical staffs and other NRC technical groups, as needed.
i.
Provides estimates of compliance burden impacts for CRGR and OMB reviews.
j.
Solicits technical revies for IEBs from designated organ-izations within the industrial sector (e.g., INP0 and AIF).
If a draft is transmitted for comment, a copy of the draft also will be sent to the PDR at the same time.
k.
Develops the related TI to provide specific guidance for R handling, evaluation, or processing of licensee replies R and onsite inspection of licensec actions.*
(All such R tis will be concurred in by DI.)
R
- To ensure that the TI is available for use by the resident inspectors and l
regional offices in a timely manner, it should be issued within one half of the time allowed for the licensees and permit holders to submit their responses (i.e., if responses are due within 30 days, then the TI should be t
issued within 15 days of the issuance of the IEB).
^
Issue Date:
02/21/86 j
- /
i NRC IE BULLETINS &-
INFORMATION NOTICES 0720-04.041 l
l 1,
Reviews the actions included in an IEB to determine the need for change in license conditions or regulatory re-quirements and requests, as appropriate, action by the appropriate licensing office or by the Office of Nuclear Regulatory Research.
I m.
Concurs in the issuance of IEBs and ins for which he/she has lead responsibility and obtains the concurrence of other NRC groups, as appropriate.
i n.
Ensures that IEBs have been properly reviewed by CRGR.
i o.
Ensures timely implementation of the action elements detailed in Section 1125-12, " Agreement on NRR/IE/AE00/-
l REGIONAL /NMSS Interface and Division of Responsibility."
j p.
Reviews all IEB responses and salient inspection reports, compiles the results, and publishes a NUREG report which gives the status of each facility, the outstanding items for followup by the regional offices and resident inspec-tors, and the recommendations for additional actions by.
Other NRC offices.
04.05 Regional Offices, t'
a.
Develops draft IEBs and ins, or request development by appropriate IE div.ision in response to certain events.and conditions within the region's purview.
h c.
Prepares and implements regional procedures to manage the i
review of responses and inspections of licensee and/or CP holder actions which contain, as a minimum, procedures for:
1.
Identification of the organizational component and/or individual responsible for review of responses.
2.
Immediate review of responses for timeliness and completeness.
3.
Verification of nonapplicability of an IEB to a site, if such is stated to be the case in the response.
k
- 'l 1
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(
1 Issue Date:
02/21/86
)
e
- ---- -. d
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NRC IE BULLETINS &
0720-04.05c.4 INFORMATION NOTICES 4.
Timely evaluation of responses for technical ade-R quacy*, if the related TI specifies regional evalu-R ation.
R 5.
Identification of the organizational component R
and/or individual responsible for inspection of R actions taken tu meet commitments made in response R to the IEB if the related TI specifies regional R inspection.
R b
7.
Documentation of completion of the various aspects of IEB closeout (items 2, 3, 4, and 6, above) in inspec-tion reports.***
d.
Ensures the periodic check and validation of the licensee's system used for the evaluation of IEBs and ins.
04.06 Resident Inspectors, a.
Identifies to the regional office any local events or conditions that, because of their generic nature, could form the basis for an IEB or IN.
- To ensure timely technical evaluation of responses, they should be com-pleted within the same time limit as that initially given the licensees to respond to the IEB (i.e., if responses are due within 30 days from the date of issuance, then the evaluation should be completed within 30 days of receipt).
For bulletins requiring emergency actions (see 03.01), the technical adequacy review should be performed as promptly as reasonable after receipt, but not necessarily with the same time span allowed the licensee.
j
- oThe timeliness of such inspection depends on the nature of the activities necessary to verify compliance with the commitments.
For instance, when j
all that is necessary is to verify that a piece of equipment has been
)
installed or modified, then the inspection should take place no later than during the next regularly scheduled inspection.
However, when it is necessary to inspect work in progress, then the inspection should be scheduled to coincide with the work activities.
QQ*It is not necessary to wait for full IEB closecut (i.e., satisfactory completion of committed activities) before reporting the status in the inspection report.
Rather, the acceptance of the earlier aspects (timeli-ness and completeness and technical adequacy) should be reported as they occur.
However, it is important that these preliminary reports clearly indicate the aspects of the IEB closecut that have been found to be acceptable.
1ssue Date:
02/21/86 _ _ _ _ _ _ _ _
h NRC IE BULLETINS &
INFORMATION NOTICES 0720-04.06b b.
Serves normally as the primary contact on matters involving licensee and permit holder responses.
- c. A larrjasyt;M,sp$Iac5ElBIUf'~$hBWfbed' i n such!1E8"CII l
ftW regiohif-d.
Checks periodically and validate the licensee's system used for the evaluation of IEBs and ins, in accordance with IE inspection program.
0720-05 CRITERIA rc c
M a.
" ' ' "? ?
'5'afeguards, or protection of the environmen.
b.
The event or condition rpust-.he fjpreys,, information, "
iridicatjiintWf5MiiiilihfEliC, pgeneric14nWture'"
'ks$
d.
Licensee response is required from addressees required to take action.
05.02 Information Notice.
An IN may be issued when events, condi-tions, or circumstances meet one or more of the following criteria:
a.
The event or condition may be important to safety, safe-guards, or protection of the environment and information is available indicating that the problem may be generic in nature.
b.
Based on the information available at the time, the event or condition does not meet the criteria for issuance of an IEB; however, licensees or permit holders should be prompt-ly notified.
c.
Information on the event or condition is preliminary in nature (essentially unevaluated by the NRC) and, when more facts are known, may result in the issuance of an IEB.
d.
There is the need to disseminate information about changes to internal NRC procedures, promulgation of new regulations I.
or modifications of existing rules and regulations that may have significant impact on NRC licensees or permit holders; this information will be disseminated only when an NRR generic letter has not been issued.
.- Issue Date:
02/21/86
D ll I
'NRC IE BULLETINS &
l 1*
0720-05.02e INFORMATION NOTICES
}
i e.
Licensee response is,n_ot, required.
0720-06 ORIGIN OF' BULLETINS AND INFORMATION NOTICES 06.01 Within IE.
Any' organizational unit within IE may recommend the issuance of an IEB or IN.
All recommendations should be pre-pared in draft form in accordance with the requirements and guidance contained herein and forwarded to the appropriate IE Division Director.
06.02 Other Sou,rces.
C.co emendations for IEBs and ins may also originate from sources other than those within IE (e.g., NRR, Regional Offices, and NHSS).
All recommendations should be prepared in draft form.in accordance with the requirements and
{
guidance contained herein and forwarded to the appropriate IE Division Director.
Section 0720-12 of this manual chapter provides the details of the IE/NRR/AE00/ REGIONAL /NMSS Interface Agreement on IEBs and ins.
j 0720-07 GUIDANCE FOR FORMAT AND CONTENT 07.01 Identification.
IEBs and ins.will be numbered using the last two digits of the calendar year followed by sequential numbers, e.g.,
83-01.
Generic communications in draft form remain unnumbered until approved for issuance.
Identification numbers for IEBs and ins are controlled by the Director of DEPER and will be placed on the document at the time of transmittal.
Each such document shall have the appropriate SSINS, OMS, and Acces-sion Numbers.
Figure 1 illustrates the use of SSINS, OMB Clearance, Accession Numbers, and identification format.
The 1
Accession Number is placed on the document by the Document Control Desk, TIDC during publication.
07.02 Format.
The format for IEBs and ins will be monitored for R compliance by DEPER.
Each type of generic correspondence R should normally have sections entitled " Addressees," "Pur-R pose," and " Description of Circumstances."
IEBs shall con-R tain a section entitled " Requested Actions."
One-or two-R page enclosures are encouraged when their incorporation serves R to clarify the information being disseminated.
R a.
~ Addressees. The addressees are to be grouped by the type of facility and license.
A typical example for an IN being sent to power rertors is:
All nucleat power reactor facilitier
..ol di ng an operating license (0L) or construction v mit (CP).
Issue Date:
02/21/86 _ _ _ _ _ _ _ _ _ _ _
e NRC IE BULLETINS &
INFORMATION NOTICES 0720-07.02a For IEBs,' the addressee section must clearly define the licensees /CP holders who are expected to take action, and those who are receiving' the IEB for information.
An example is:
Addressees:
For Action:
All pressurized water nuclear power reactor l
facilities (PWRs) holding an operating license (OL).
For Information:
All other nuclear power reactor facilities holding an OL or construction permit (CP).
b.
Purpose. The purpose of the IN or IEB should be summarized in this section of the IN or IEB.
An example for an IN is:
This information notice is' provided to alert recipi-ents of a potentially significant problem pertaining j-to (applicable component, system, or subject)..It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem occurring at their facilities.
However, suggestions contained in this notice' do not constitute-NRC re-quirements; therefore no specific action or written response is required.
Occasionally, the above wording may have to be modified to agree with the subject of the information notice, e.g.,
notification of a
rule change which will require compliance.
An example for an IEB' is:
The purpose of this bulletin is to:
(1) notify licensees and construction permit holders about l
incidents of severe degradation of threaded fasteners (bolts and studs) in closures in the reactor coolant pressure boundary, and (2) to -request appropriate actions.
(
! Issue Date:
02/21/86 l
i I
NRC IE BULLETINS &
0720-07.02c INFORMATION NOTICES c.
Description of Circumstances.
The
- problem, event, or circumstances must be clearly described to the extent that available information will permit.
The description must be factual and brief.
Where the licensee or permit holder has identified or has assigned a cause to the event, that j
information should be provided, including the source from which the information was obtained.
Equipment manufacturers and model numbers should be identi-fied in the description section whenever such information i
is critical to the communication and is important in the identification and resolution of the problem.
Except where confidentiality may be involved, the original source of information on which the IEB or IN is based (e.g., a licensee, vendor, permit holder, applicant, etc. )
should be identified.
Although not encouraged, opinion and speculation statements may be included when attributed to such a source.
Remedial or corrective actions taken to cope with the problem may be included provided such actions are:
(1) attributed to the source of information and deemed to be reasonably correct by NRC, or (2) NRC specifi-cally identifies and requests that such corrective action be taken (IEBs only).
L d.
Request for Information/ Action - Bulletins.
Information needs should be very specific and limited to that deemed necessary.
The information must not be available elsewhere in a reasonable time.
Insofar as practicable, information should be requested in a form amenable to quick evaluation.
Short, numbered paragraphs should be used to treat each separate element of the request.
Where beneficial, data sheets or checklists should be considered and used to expedite information collection by the licensee and evalua-tion upon receipt by the NRC.
This section must contain the following:
1.
The category,' group, or class of licensees / permit holders / applicants / vendors affected by the request.
j 2.
The specific action (s) requested.
1 3.
The date(s) when action and/or response is due.
l Action statements in IEBs must be reviewed to ensure that l
l they do not impose any continuing requirements.
IEBs are j
not intended, and must not be used, to substitute for new l
(or revised) license conditions or other continuing l
l requirements i
1
.?
1 Issue Date:
02/21/86 _-_
______-_________ ~
i NRC IE BULLETINS &
INFORMATION NOTICES 0720-07.02e f
e.
Closure - Bulletins.
All IEBs issued to licensees and applicants must include the following statements in their j
closure:
The written reports requested by items and 1
should be submitted to the appropriate Regional Administrator, signed under oath or affirmation, under
)
the provisions of Section 182a of the Atomic Energy Act of 1954, as amended.
In addition, the original of the cover letters and a copy of the reports shall be transmitted to the U.S. Nuclear Regulatory Commission, j
Document Control Desk, Washington, D. C.
20555 for reproduction and distribution.
Although the specific details involving (the applica-f ble systems are to be delineated by the author) may j
not directly apply for your' facility, you are request-ed to review the general concerns expressed in the bulletin for applicability. at your facility.
Your response should describe the results of the review, and if the general concerns apply, you should describe the short-term and long-term corrective actions to be taken and the schedules thereof.
4 The wording of the previous paragraph may have to be modified to reflect the subject and intent of the IEB.
j
- However, as a minimum, such statements should require y
negative responses from all addressees required to take action to ensure that they have received and evaluated the j
IEB.
I Next, the following statement on OMB clearance is to be included:
This request for information was approved by the Office of Management and Budget under clearance number (see Section 0720-13).
Comments on burden and dupli-cation should be directed to the Office of Management and
- Budget, Reports Management, Room 3208, New Executive Of fice Building, Washington, D.C.
20503.
j The following statement should be used as the final para-graph of the IEB:
If you have any questions about this matter, please l
contact the Regional Administrator of the appropriate NRC regional office, or the technical contact listed
- below, f.
Closure - Information Notices.
The following statements
(
are to be included after the " Description of Circumstances" section: Issue Date:
02/21/86
ny
- ..e NRC IE BULLETINS &
- 0720-07.02f INFORMATION NOTICES.
No ' specific action or wdtten response is required by
'this information notice.
If you have any - questions si about this matter, please contact the Regional Admin-istrator of the appropriate NRC regional office or this office.
- g..
Technical Contact.
The name of a technical contact and phone number should be included.
07.03
_Recently Issued IEBs or ins.
A list of the most recently issued
. IEBs or ins will be attached to the final document package.
0720-08 DISSEMINATION AND DISTRIBUTION PROCEDURE
'0B.01 Advance Copies.
Following headquarters concurrence and approv-al, advance copies of the IEB or IN will be transmitted to the i
regions and to the originating IE branch by electronic means or 1
other suitable means, as appropriate.
~
i 08.02 IE Headquarters Distribution.
At the' time an advance copy of R the IEB or IN package is transmitted, DEPER will arrange. for R distribution to internal and external addressees as specified R l
on standard distribution lists through the DMB.
R Additionally for IEBs, DEPER will immediately send copies to R the DMB.
R 0720-09 REPLIES TO BULLETIN - ACTIONS 09.01 Distribution.
Replies to IEBs will be distributed by the DMB.
09.02 Adequate Replies.
Each fccility's response should be evaluated for timeliness, completeness, and technical adequacy in accor-dance - with regional procedures.*
For those replies that indi-cate adequate action and response by the licensee or permit holder and where the corrective action is not complicated, the matter should be included as a followup item during a subsequent inspection for closeout and resolution in accordance with regional procedures'.
If prompt resident inspector or regional inspection and/or
)
followup are required, as may be the case with some IEBs, instructions will be provided by the cognizant headquarters division in the related TI.
- In some cases special evaluation instructions will be issued that may require establishing regional and/or headquarters evaluation teams.
Such a
cases will be identified by the cognizant headquarters division in the related TI.
.t 1ssue Date:
02/21/86 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ -
NRC IE BULLETINS &
INFORMATION NOTICES 0720-09.03 09.03 Inadequate Replies.
If a reply is dete 71ned to be inadequate,R the region should request the licensee to provide an adequate R response.
The appropriate IE Division may in pre-determined R cases inform the licensee directly in those cases evaluated by R IE.
If a licensee is unwilling to provide an adequate solu-R tion or to demonstrate that the action is not required, the R
)
matter must be brought to the attention of headquarters for R consideration and possible issuance of an NRC Order.
R 09.04 Weplies Involving Alteration of Existing License Conditions.
If the reply f rom a licensee or permit holder includes actions that appear to involve changes to the license, technical specifica-tions, or SAR, or if a topic involves an unreviewed safety
- question, the region should inform the appropriate project manager in NRR's Division of Licensing in accordance with the NRR/IE/AE0D/ REGIONAL /NMSS Interface Agreement (Section 0720-12).
09.05 Replies to IEBs normally are not acknowledged.
09.06 Followup NUREG Report.
Following review of all IEB responses and salient inspection reports, the IE Branch having lead responsibility will compile the results and publish a NUREG report or memorandum which gives the status of each facility, outstanding items for followup by the regional offices, and recommendations for additional actions by other NRC of fices.
0720-10 REVISIONS AND SUPPLEMENTS TO BULLETINS AND INFORMATION NOTICES.
Efforts will be made to provide full and sufficient information in the initial issue of all IEBs and ins.
However, because of time constraints, it may be necessary to promptly issue prelimi-i nary information and then provide additional information as it
]
becomes available.
In such cases, any revisions or supplements
{
will be processed generally as describ'ed herein for initial issuance and will be identified using the original number followed by a revision or supplement number.
If after being issued, the original generic correspondence reovires correction because the preliminary information was incorrect, this subsequent information will be issued as a revision and will supersede the original document.
If, however, the original generic correspondence requires updating to provide new information or change a listing of affected licensees (listed in the text and not as addressees), this type of addi-i tional information will be issued as a supplement.
Changing addressees, e.g., class of licensees or major change in content, j
necessitates issuing a new generic communication with a separate unique number.
In any case, the reason (s) for issuing another
(
generic correspondence on the same subject is to be addressed l
specifically in the " Purpose" section of the document. Issue Date:
02/21/86
___________-----.-----_______J
NRC IE BULLETINS &
0720-11 INFORMATION NOTICES 0720-11 INFORMING NEW LICENSEES AND PERMIT HOLDERS To inform prospective licensees about IE generic communications, the regions 'may select a representative sample of ins and IEBs issued previously and provide these examples to new licensees and permit. holders at the first Regional / Corporate Management meeting.
0720-12 A'GREEMENT ON NRR/IE/AE0D/ REGIONAL /NMSS INTERFACE AND DIVISION OF RESPONSIBILITY
.2 has the responsibility for issuing IEBs and ins.
IEBs are i
issued when a significant safety or safeguards issue is in-volved, when prompt involvement of licensees is desired, when specific actions are requested from the licensee, or when a q
response is requested from the licensee.
ins are issued to j
provide prompt information to licensees, perhaps even before all j
facts are known.
An IN may precede the issuance of an IEB.
)
will consider issuing IEBs and ins that may be proposed by any i
NRC office, but retains ultimate responsibility for the decision regarding issuance.
If time permits, IE will give the regional offices, AE00, NRR and/or NMSS an opportunity to comment on all proposed IEBs pertaining to licensees before their issuance.
Also, any office providing a proposed IEB or IN will be requested to review it before final issuance by IE.
IE will respond to significant comments not incorporated in an IEB.
When a proposed IEB requests action that NRR or NMSS declares to be in potential conflict with existing license requirements, or if for other valid reasons a specific request requires that a proposed IEB not be issued, the matter will be resolved at the Division Director level or higher as appropriate.
If IE has the responsibility to evaluate licensee responses to IEBs, IE will provide a summary of the responses, if action is requested.
All, responses, together with the summary and possi-ble recommendations for resolution, will be forwarded to the
{
Regional Offices, AE0D, NRR, and/or NMSS.
In addition to the normal evaluations in accordance with region-al procedures and the alternate evaluations by IE, evaluations may be performed by NRR and/or NMSS.
Also, certain issues may require combining resources from two or more of the aforemen-tioned offices to ensure a comprehensive evaluation.
Resource commitments will be made at the Division Director level.
Such cases will be identified by the cognizant headquarters division in the related TI.
i 3
Issue Date:
02/21/86 ___
J
-.-----..m-7 4
+,
NRC IE BULLETINS &
0720-13 INFORMATION NOTICES 0720-13 OMB CLEARANCE FOR BULLETINS Under the Paperwork Reduction Act (PRA) of 1980, Public ' Law 96-511, the OMB is required to review the collection of informa-tion by independent regulatory agencies to ensure that the information required by the agencies is obtained with a minimum of burden on respondents and does not duplicate information already collected by Federal agencies.
In addition, the inde-
)
pendent regulatory agencies shall not collect identical informa-tion from 10 or more respondents unless advance approval has been'obtained from the OMB.
To carry out its mandate, OMB promulgated rules and regulations, requiring the following:
a.
Each agency to submit all proposals for the collection of
!dentical information or record keeping requirements from 10 or more persons for advance approval.
b.
Each agency to plan and conduct its information collection activities, including establishing procedures for managing such activities, in a manner consistent with. the PRA.
The requirement for advance OMB approval for each information request could be contrary to the NRC mandate of protecting the health and safety of the public (42 U.S.C. 2012[d]),-especially where an event at one licensee facility strongly suggests a nonroutine generic safety problem at other licensee plants.
Accordingly, so that the NRC is not unduly restricted, OMB 'has granted to the NRC preapproved clearance numbers for use in issuing bulletins.
For instances where CRGR review and approval are not obtained because of the emergency nature of the issue, OMB clearance number 3150-0012 shall be used.
For instances where CRGR review and approval are obtained before issuance of an IEB to reactor licensees, vendors, or construc-tion plant holders, OMB clearance number 3150-0011 shall be used.
In all other instances where an IEB will affect 10 or more public entities, the Administrative Section, PRSB will obtain the appropriate OMB clearance and provide the OMB number, i
END
( Issue Date:
02/21/86
)
NRC IE BULLETINS &
JNFORMATIONNOTICES
- FIGURE 1, 0720 SSINS No.:
6820 OMB No.:
3150-0011 IEB:
82-04 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 December 3, 1982
.IE BULLETIN NO. 82-04:
DEFICIENCIES IN PRIMARY CONTAINMENT ELECTRICAL PENETRATION ASSEMBLIES Addressees:
All nuclear power reactor facilities holding an operating license (OL) or construction permit (CP).
No. 80-09 associated with electrical penetration assemblies supplied at.
Bunker Ramo. The deficiencies included improperly crimped lugs and improp-erly identified penetration cables.
During hand pull tests, at least.38 wires separated from their lugs.
It was reported that this deficiency l
resulted when Bunker Ramo overcrimped and undercrimped lugs.
i 8208190259 Accession Number * (inwer left corner)
(a)
SSINS No.:
6835 IN 82-02 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 January 27, 1982 IE INFORMATION NOTICE N0. 82-02:
WESTINGHOUSE NBFD RELAY FAILURES IN REACTOR PROTECTION SYSTEMS AT CERTAIN l
NUCLEAR POWER PLANTS (b)
Figure 1:
Identification Format for IEBs, and ins (SSINs, OMB Clearance, and Accession Number usage)
(a) bulletins, (b) information notices i
I
- Accession number location:
identical for ins,
F-1 Issue Date:
02/21/86
_ _ ______ - --____