ML20129J744
| ML20129J744 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/29/1996 |
| From: | Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | James Knubel GENERAL PUBLIC UTILITIES CORP. |
| References | |
| EA-95-238, NUDOCS 9611080023 | |
| Download: ML20129J744 (3) | |
See also: IR 05000289/1995016
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' October 29, 1996
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EA 95-238
Mr. James Knubel
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Vice President and Director, TMl
GPU Nuclear Corporation
Three Mile Island Nuclear Station
Post Office Box 480
Middletown; Pennsylvania 17057-0191
SUBJECT:
INSPECTION REPORT 50 289/95-16 REPLY
Dear Mr. 'Knubel:
This' refers to your April 10,1996, correspondence in response to our letter, dated
March 11,1996, regarding the Three Mile Island Nuclear Station, Unit 1 (TMI-1). - This
- correspondence included your response to the two violations delineated in the subject
inspection report. The violations dealt with your actions in addressing past problems with
pipe supports on the reactor coolant system (RCS) drain lines. We have reviewed this
matter in accordance with NRC inspection Manual Procedure 92903, " Engineering."
With regard to the first violation, we concur with your root cause assessment. We
understand the root cause is attributed to the failure to capture the recommended pipe
support modification in a formal tracking process; failure of the GPUN staff and
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management to followup on the issue; and personnel errors in performing design
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calculations and design verifications. Further, we understand that the seven corrective
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actions described in your letter have been completed, and we concur that these actions
provide reasonable assurance that a similar event will not occur in the future.
With regard to the second violation, we concur with part of your assessment, and
associated corrective actions. Specifically, we understand the root cause of failing to meet
the requirements of ASME Section !!! d attributed to a lack of understanding on the part of
the individuals performing the design calculation and verification. Further, we understand
that the five corrective actions described in your April 10 letter have been completed, and
we concur that these actions provide reasonable assurance that a similar event involving
failure to meet ASME Section lli criteria will not occur in the future.
We disagree with part of your response to the second violation in which you suggest that
the use of engineering judgement to evaluate nonconforming conditions, in lieu of specific
ASME code requirements, is acceptable and within the guidance of ASME Section XI. We
note that your current code of record,1986 ASME Section XI, subsection IWF-3410,
provides acceptance standards for inspecting component supports, and states that a
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Mr. James Knubel
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deformed pipe support is an unacceptable condition requiring repair, replacement or
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analysis. Your position, as stated in your April 10,1996, letter, suggests that engineering
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Judgement can be used in lieu of these acceptance standards for dispositioning deformed
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supports. We disagree with that position.
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In your letter, you state that the use of engineering judgement to evaluate existing
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conditions is acceptable and within the guidance of the 1986 ASME Section XI code. Your
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letter does not specify where such guidance, allowing the use of engineering judgement in
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lieu of satisfying code acceptance criteria,is specified in the ASME code. We are
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concerned that your application of engineering judgement in lieu of code requirements
could lead to situations similar to the RCS drain line in which nonconforming conditions are
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incorrectly dispositioned and returned to service with no corrective actions. Consequently,
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please provide within 30 days of this letter, a specific reference to the section of ASME
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Section XI that allows the use of engineering judgement in lieu of code specified
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ecceptance criteria, and an explanation of how your current ASME Section XI
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implementation program precludes situations similar to the RCS drain line.
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Sincerely,
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James T. Wiggins, Director
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Division of Reactor Safety
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Docket No. 50-289
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License No. DPR-50
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cc w/ encl:
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E. L. Blake, Shaw, Pittman, Potts and Trowbridge (Legal Counsel for GPUN)
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Commonwealth of Pennsylvania
J. C. Fornicola, Director, Licensing and Regulatory Affairs
M. J. Ross, Director, Operations and Maintenance
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TMI-Alert (TMIA)
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J. S. Wetmore, Manager, TMI Licensing Department
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Mr. James Knubel
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Distribution w/ encl:
Region I Docket Room (with concurrences)
.T. Kenny, DRS
P. Eselgroth, DRP
D. Haverkamp, DRP
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NRC Resident Inspector
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Nuclear Safety Information Center (NSIC)
PUBLIC
W.-Dean, OEDO
Inspection Program Branch, NRR (IPAS)
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DRS File (1)
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DOCUMENT NAME: A: TMI9616. REP
To rwelve a copy of this document, indicate in the box: "C" = Coty}vithout attachment / enclosure
"E" = Copy with attachment / enclosure
'N' = No copy
0FFICE
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DATE-
05/31/96
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0FFICIAL RECORD COPY
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GPU Nuclear Corporation
Nuclear
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Middletown. Pennsylvania 17057-0480
(717)944-7621
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Writer's Direct Dial Nurnber:
(717) 948-8005
6710-96-2024
April 10,1996
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U. S. Nuclear Regulatory Commission
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Attn: Document Control Desk
Washington, D. C. 20555
Gentlemen:
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Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
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Operating Ilcense No. DPR-50
Docket No. 50-289
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Response to Notice of Violation
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Reference:
NRC Letter dated March 11,1996, " Notice of Violation (NRC
Inspection Report No. 50289/95-16)".
'Ihe referenced letter enclosed a Notice of Violation containing two violations and states that
...the first violation involved GPUN's failure to control ad=>ately a modification to the
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RCS drain line piping" and the "...second violation occurred when GPUN performed a
more refined ASME Section III *1atian to disposition the pipe overstress condition
calculated in the 1990 B31.1 analysis".
Pursuant to the provisions of 10 CFR 2.201, Attachment I to this letter provides the GPU
Nuclear response to each of the two violations contamed in the Notice of Violation.
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Sincerely,
m
J. Knubel
1/ ice President and Director, TMI
RTZ
cc:
TMI Senior ResidentInWor
TMI-1 Senior Project Manager
. Region Adm' istrator
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GPU Nuclear Corporation is a subsidiary of General Public Utilities Corporation
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METROPOUTAN EDISON COMPANY
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JERSEY CENTRAL {0WER AND UGHT COMPANY
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PENNSYLVANIA ELECTRICAL COMPANY
GPU NUCLEAR CORPORATION
Three Mile Island Nuclear Station, Unit 1 (TMI-1) -
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Operating Ucense No. DPR-50
Docket No. 50-289
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Reply to Notice of Violation (NRC Inan~* inn _ Report No. 50-289/95-16)
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tanamie'~i by NRC Imer dated March 11,1996.
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This letter is submitted in reply to the Notice of Violation (NRC Taan~+ inn Report No.
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50-289/95-16) transmierari by NRC I.etter dated March 11,1996 which refers to the
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inspection conducted on September 25-29,1995, at the 'Ihree Mile Island, Unit 1 Nuclear
' Station (IMI) facility and from October 10-11,1995, at the GPU Nuclear (GPUN) Office
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in Parsippany, New Jersey. All emeements containari in this reply have been reviewed, and
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all such mataments made and matters set forth therein are true and conect to the best of my
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knowledge.
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Vice President and Director, TMI-1
Signed and sworn before me this
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day of
1996
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f.c f il : a;
Suzarne ::. M kuh, N:.ter/PuMic
Londondctry Tv@., Carnin Coun*y
My Comrritt on Expires Nov. 22,1999
Memter Penrr,Mr.u Associanon orwuu.s
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'6710-96-2123
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Attachment 1
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Page 1 of 6
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ATTACHMENT I
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A.
10 CFR Part 50, Appendix B, ' Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants,"Section III, tesign Control," requires,
in part, that measures be ectablished for the identification and control of d
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interfaces and for coordination among participating design organizations.
Appendix B also requires, in part, that licensees establish design control
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measures that provide for verifying or checking the design adequacy.
Contrary to tne above, from 1990 through at least September 1995, measures
were not established for the identification and control of design interfaces and
for coordination among participating design organizations as required, nor did
the licensee establish design control measures that provided for verifying or
checking design adequacy.
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1. Specifically, GPUN failed to control adequately a modification to the
Reactor Coolant System (RCS) drain line piping. 'Ibe modification was
developed as a result of a 1990 GPUN structural analysis that demonstrated
that the drain line piping was oventressed due to an improper support
configuration. 'Ihe modification was described in a letter, dated August 27,
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1990, transmitted from GPUN Headquarters in Parsippany, New Jersey to
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the TMI site. However, the modification was not implemented as of
September 1995, and GPUN could provide no documentation to demonstrate
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that the modification was ever properly disresitioned.
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2. GPUN's design verification process failed to identify a significant error in
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the 1990 analysis that resulted in underestimating the level of stress in the
pipe. Specifically, the analysis indicated thennal expansion stresses
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approximately 4% above the allowables specified in the design code of
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record (USAS B31.1 1%7). However, when the analytical error was
corrected in 1995, the stresses were approximately 100% above the code
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allowables. (01013).
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This is a Severity Level III violation (Supplement I).
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6710-96-2123
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Attachment 1
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GPU NUCr RAR BRCPONSE TO THE NOTICE OF VIOLATION (Viointian Al
GPU Nuclear acknowledges that the violation occurred as stated in the notice of
violation. The first part of the violadon, i.e., Part 1, occurred because the
recommendation was not captured in any formal tracking system and there was no
follow-up action on the part of staff or cog tizant management.
GPU Nuclaar ham imniamantad the followine corrective actions for Part 1 of this
violatian:
1)
GPU Nuclear has implemented a modification to the RCS drain line supports
which satisfies B31.1 Code requirements.
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All individuals involved were counseled that a lack of follow up action leading to
.a proper disposition of the recommendation is not acceptable.
3)
Since 1990, the Project Approval and Management Process has been
reengineered. Specifically, the process today requires that the responsible System
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Engineer be involved with all proposed modifications, and that the System
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Performance Team reviews all modifications involving multi-diciplinary reviews
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where and when appropriate.
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Management has emphasized to the staff that proper close out of
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recommendations is an essential part of the engineering process.
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With respect to,the practice of proposing modifications in informal
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communication systems, GPU Nuclear 1.as conducted a search of documents
generated by the E=p= ring sul Design Department where such
recommendations may have existed. In particular, all memoranda and Technical
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Data Reports from the years 1988,1990 and 1993 were reviewed, as well as
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Technical Functions Assigned Action Items from 1990. In all, over 4,000
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documents were reviewed. It has been cencluded from this review that the
specific incident identified in the violation is an isolated case.
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All the above actions have been completed as of the date of this response. GPU
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Nuclear believes the actions taken provide reasonable assurance that a similar event will
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not occur in the future.
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6710- % -2123
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Attachment i
Page 3 of 6
The second part of the vietation, i.e. Part 2, occurred due to personnel errors on the part
of the preparer and verifier of the calctlation. The Mechanical Analysis Staff within GPU
Nuclear, which performs the pipe stress analyses, is located in fM corporate offices in
Parsippany, NJ. In 1995 the group consisted of five engineers with an average of 20 years
nuclear experience. Four of the five engineers hold Masters Degrees and four hold a
Professional Engineer License. 'Ihe engineer who prepared the calculation and the
engineer who verified the calculation are and were technically qualified to perform their
functions.
GPU Nuclear identified this error during its investigation of the crack in the drain line
during the Three Mile Island Unit 1 Cycle 11 Refueling (11R) outage in 1995. As
stated in the violation, when the specific error was corrected the calculated stresses
were approximately 100% above the allowable specified in the design code of record
(USAS B31.1 1%7). That was n interim result. Further analyses, which more
accurately represent the actual conditions for the pipe, conclude that the stresses were
40% above the allowable. 'Ihese analyses also demonstrate that these stress conditions
were not contributory to the initiation or propagation of the crack. 'Ihe modifications
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made during the last refueling outage have reduced these stresses to values well below
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Code allowable.
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GPU Nnetaar kna imniamaatad the followina corrective n_tions for Part 2 of thic
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viointian:
(1)
The individuals involved were counseled; the verifier was temporarily restricted
from performing verifications until he was retrained on the enhancad design
verification procedure; and management has emphasized that both the preparer
and verifier of a calculation must perform their functions with appropriate focus
on technical quality. Further, a sample of previous verifications perfonned by
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the verifier was reviewed. It has been concluded that the incident noted in the
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violation is isolated.
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(2)
GPU Nuclear has reviewed the governing procedures in an effort to identify
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programmatic issues that may have contributed to the cause of the calculation
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and verification error. Further, GPU Nuclear consulted with others in the
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industry including two utilities and two architect engineering companies to learn
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about other design verification progams and practices. Although GPU Nuclear
has concluded that our existing procediaes did not contribute to the cause of the
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incident cited in the violation, these procedures were updated and anhancad to
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provide more complete guidance and direction. 'Ihe corporate headquarters
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engineering staff, including management, involved with design calculations and
verification were retrained on the procedure enhancements.
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All the above actions have been completed as of the date of this response. GPU
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Nuclear believes the actions taken provide reasonable assurance that a similar event will
not occtirin the future.
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6710 96-2123
Attachment 1
Page 4 of 6
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NQTICE OF VIOLATION
(Viohtion B)
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10 CFR 50.55a, ' Codes and Standards", paragraph (g), ' Inservice inspection
requirements," requires that licensees of nuclear power plants meet applicable
criteria in Section XI, Division 1, of the American Society of Mechanical
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Engineers (ASME) Boiler and Pressure Vessel Code ' Rules for Inservice
. Inspection of Nuclear Power Plant Components." ASME Section XI requires,
in part, that licensees perform inservice examinations of class I compor,ents,
including supports. If the components or suppons do not meet the examination
acceptance criteria,Section XI requires that the licensee perform a repair or
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replacement, or perform further evaluation to demonstrate the adequacy of the
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components or supports.
Contrary to the above, in 1988 and 1990, GPUN did not meet the applicable
requirements in 10 CFR 50.55a, in that, during the performance of inservice
inspection (ISI) examinations of pipe supports on the RCS drain lines, a class I
component, GPUN identified distorted snorts and failed to repair or replace
or perform an adequate evaluation to cuablish the adequacy of the piping and
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supports. GPUN performed a structural analysis of the drain lines that
demonstrated that the configuration of the supports did not allow for adequate
pipe thermal expansion. Consequently, stress levels in the drain line piping
exceeded the allowable stress values specified in the piping design code of
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record, USAS B31.1-1%7, ' Power Piping". GPUN utilized a later code,
Section III subsection NB-3653.6, to disposition the overstressed piping.
Paragraph NCA-1140 of Section III allows the use of specific provisions of the
code but requires that all related requirements be met. The analysis performed
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by GPUN did not incorporate all related requirements of NB-3653.6 and was
inadequate to demonstrate the adequacy of the piping and associated supports.
(02014).
This is a Severity LevelIV violation (Supplement I).
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Attachment 1
Page 5 of 6
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GPU NUCr RAR RRCPONSE TO T5tR NOTICE OF VIOLATION (Viointian B)
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GPU Nuclear acknowledges that the violation occurred as stated in the notice of
violation. This violation consists of two parts, one relates to support inspections and
evaluations performed in 1988 and 1990, and one relates to a pipe stress calculation
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performed in 1990. The supports were evaluated under the ASME Section XI 1977
Code through the 1978 Summer Addenda. 'Ihe accynace standards portion of this
code, Section IWF 3410, was in the course of preparation. Therefore, Engineering
was requested to determine if the support configuration was acceptable for continued
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service or required repair, replacement or further evaluation. Engineering determined
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that since the observed distortions were minor, the supponts were accymble as is and
'would perform their intended function. 'Ihe engineer used his judgment based upon his
knowledge of materials and the design conditions to arrive at his conclusion. GPU
Nuclear maintains that the use of engineering judgment to evaluate existing condiuces
is acceptable and within the guidance of ASME Section XI 1978 and 1986 (the ::urrent
version for TMI).
'Ihe pipe stress analysis issue relates to the full use of the appropriate guidance in
ASME III when evaluating an issue from a B31.1 analysis. GPU Nuclear agrees that
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not all related requirercents were satisfied. This situation occurred because both the
preparer and verifier were not fully aware of the relationship between the requirements
of ASME III and B31.1.
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As a matter of perspective, the applied number of thermal cycles for this pipe is 240.
If the B31.1 allowable stress had been met the allowable number of cycles per B31.1
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would have been 7000. The overstress that the engineer was attempting to resolve was
4% beyond the B31.1 code allowable. 'Iherefore it was his conclusion before he began
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the rtconciliation, that the stresses would be acceptable because the difference in
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applied stress vs. allowable stress was so small.
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GPU Nuclear has implemented the following corrective actions for this violation:
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1. The engineers involved were counseled that it is imperative that when issues are
dispositioned, the technical logic leading to conclusions must be clearly
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documented.
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2. Engineering and Design department management has emphasized to engineers that
adherence to all relevant code provisions is required.
3. An outside consultant was retained to investigate and explain the issues in applying
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ASME III when evaluating a B31.1 analysis.
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6710-96-2123
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Attachment 1
Page 6 of 6
4. 'Ihe calculation procedure has been revised to ensure management concurrence is
obtained when specialimi methods are used to satisfy the design code of record,
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i.e., ASME Section M or B31.1 evaluations. Management has stressed that
adherence to the design code of record is a design basis requirement.
5. GPU Nuclear reviewed the stress analysis files and found no other instance where
ASME M was used to reconcile design basis issues.
All the above actions have been completed. GPU Nuclear believes the actions taken
provide reasonable assurance that a sunilar event will not occur in the future.
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