IR 05000237/1993005
| ML17179A712 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/27/1993 |
| From: | Beverly Clayton, Hiland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17179A711 | List: |
| References | |
| 50-237-93-05-EC, 50-237-93-5-EC, EA-92-253, NUDOCS 9302100432 | |
| Download: ML17179A712 (18) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION REGION II°!
Report No; 50-237/93005(DRP)
Docket N License No. DPR-19 Licensee:
Commonwealth Edison Company Executive Towers West III 1400 Opus Place, Suite 300 Downets Grove, IL 60515 Meeting Conducted: *January" 22, 1993 *
Meeting Location:
Region III Office 799 Roosevelt Road Glen Ellyn, Illinois 60137 Type of Meeting:
Enforcement Conference In$pection Conducted:
Dresden Site, Morris IL October 19 to December 15, 1992 Inspecto~:. Pat~{ci~ Lougheed Reviewed by:
Approved by:*
Patrick L. Hilan Reactor Projects Brent Clayton, CHief Reactor Projects Branch 1 Meeting Summary:
IB
'
1{.,4;
-~*
1/17/.PJ Date EA 92-253 Enforcement Conference on January 22. 1993, (Report No. 50-237/93005{DRP))
A~eas Discussed:
An apparent violation identified du~ing the inspection was discussed, along with the corrective actions taken or planned by the license The enfo~ce~ent options pertaining to the apparent viol~tion were also discussed with the litensee.. The ap~arent violation concerned a Unit 2 Technical Specification revisi-0n which contained_ inaccurate information when submitted to the NR I,.
9302100432 930128
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-PDR ADOCK *osooo237-- --- ---
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- DETAILS.
Persons Present at the Conference Commonwealth Edison Company L. DelGeorge M. Lyster H. Massin D. Shamblin R. Ward D. Jenkins S. Truhatch P. Barnes S. Eldridge J. Shields D. Taylor B. Adams P. Ptet W. Morgan s~ Reese-:-Koenig D. Saccomando T. Spry L. Taylor E.. White Vice President, Nuclear Over~ight and Regulatory Services Dresden Station Vice President Site Engineering and Construction Manager Engineering and Construction (ENC) Man~ger Regulatory Performance Manger Counsel, Commonwealth Edison Company Counsel, Winston and Strawm
- Compliance Supervisor Dresden Engineering Plant Support Supervisor Dresden Regulatory Assurance Supervisor
- ENC Regulatory Assurance Supervisor ENC Regulatory Assurance Supervisor Dresden Nuclear Licensing Administrator Dresden Station Partner Compliance Engineer Compliance Engineer Welding and Materials Engineer ENC Regulatory Assurance Engineer ENC Regulatory Assurance Engineer U. S. Nuclear Regulatory Commission A. Bert Davis R. DeFayette B. Berson P. Hiland P. Pelke M. Leach M. Peck D. Jones
- B. Siegel*
W.* Troskoski j, Tsao Regional Administrator, Riii
- Director, Enforcement and Investigations Staff, Rill Regional Counsel, Riii
- *
Chief, Reactor Projects Section lB, Rill Enforcement Specialist, Rill Senior Resident Inspector, Rill Resident Inspector, Riii Reactor Engineer, Riii Dresden Project Manager, Office of Nuclear Reactor-Regulation (NRR)
Enforcement Specialist, Office of Enforcement Engineer, Materials and Chemical Engineering Branch, NRR
.
- Enforc~ment Conference An enforcement conference was held in the NRC Region III office on January 22, 199 This conference was conducted as a result of the preliminary findings of the inspection conducted on October 19 to December 15, 1992, in which an apparent violation of NRC regulations
was identifie Inspection findings were documented in Inspection Report 50-237/92035(DRP), transmitted to the licensee by letter dated December 24, 199 *
The purpose of this conference was tq {l) discuss the -apparJnt
- viol~tion, its causes, and the licensee's corrective actioris; (2) determine if _there* were any escalatjng or mitigating circumstances; and (3) obtain any informatiori which would help determine the appropriate enforcement actio.
The licensee's re~resentatives provided additional information concerning the apparent violatio The licensee's representat.ives
.described the events which led to the apparent violation, including root causes and corrective actions taken.. The licensee's presentation slides
. are provide~ as Attachment l to this repor At the conclusion of the meeting, the licen~ee was informed that th~y would be notified in the near future of the final enforcement actio Attachments: CECo Pr~sentation Slides 3 -
JANUARY 22, 1993 ENFORCEMENT CONFERENCE DRESDEN STATION.
.
.
.
. ACCURACY OF INFORMATION TO. SUPPORT TECH SPEC AMENDMENT AGENDA
..
INTRODUCTION/OVERVIEW MICHAEL LYSTER EVENT CHRONOl,..OGY TOM SPRY.*
/,.
,,
SAFETY SIGNIFICANCE TOM SPRY r
h I CORRECTIVE ACTIONS HERB MASSIN i1' r.:.
CONCLUSIONS MICHAEL L YSTE:-t lNLD/2458-1
- .
OVERVIEW OF EVENT VIOLATION Attachment*!
- . Page 2 of 15
- .. CECo acknowledges that in 1989 inaccurate information was inadvertently used
. to support a Dresden Unit 2 Technical Spec~ication Amendment Reques APPARENT CAUSES In determining the impact of Generic Letter.88-11, "Radiation Embrittlement of
.Reactor Vessel Material," reactor vessel fabrication summary data was reviewed. The data was misinterpreted which led to an ~erroneous report conclusion. The error was not identified during the supervisory review-and the report subsequently was used in the development of new Tech Spec Pressure Temperature curves;
/
SAFETY SIGNIFICANCE Reactor vessel integrity was not impacted by this event. Total margin against.
crack initiation exceeded minimum ASME allowable limits. Had the incorrect PT curve remained in effect, there would have been no impact on vessel integrit CORRECTIVE ACTION Even though this situation_ was not discovered until 1992, CE Co recognized in 1989, that actions should be taken to improve the processes for the performance, review, and approval of design calculations done by both CE Co and it's *.
contractors. Sever~I new engineering programs and procedures have been.
- developed ~nd enhanced to ensure the quality of engineering document ZNLD/2458::2 ________ _: ______ --.
04/85 07/12/88" EVENT CHRONOLOGY Attachment 1 Page 3 of 15 CECo purchased from GE a summary of reactor vessekmaterial *
and fabrication information for the B& W fabricated Dres~en and.
. Quad-Cities reacior pressure vessels (BPVs):
Generic Letter 88-11, "Radiation Embrittlement of Reactor Vessel Materials," was issued. The Generic Letter.required licensees to
- evaluate Pressure-Temperature (PT) Curves against Beg. Guide 1.99 Bev. 2. "Radiation Em.brittlement of Reactor Vessel Materials."
Evaluation of the impact of Generic Letter 88-11 and Beg. Guide 1.99 Bev. 2 on.Tech Spec PT curves was assig-ned to a loanee in the ISi & Materials Grou The loanee, an employee of EPBI and the manager of the BPV Embrittlement Management Progra.m was a recognized-industry expert. He was loaned to CECo to gain an understanding of the.
CECo organization and processes. * He acted as a CECo employee during his time in the Production Services Department ISi &
Materials Grou.
.
10/19/88 The Generic Letter 88-11 evaluation was completed, utilizing the *
1985 GE Summary Data Tables which were based on B&W
- .
ZNLD/2458-3 fabrication records. A report was written with recommendations for determining pressure-temperature limits. It was concluded that the..
Dresden 2 reactor vessel could be considered separately from the other units (D3, 01, 02) based on a determination that the limiting material in the ves$el was not the electroslag weld. This conclusion was subsequently determined to be in erro Summary Data Table information was unclear *(see Attachments 1 and 2).
.
.
... *
_.
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_
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Two weld procedure qualification r~cord numbers were listed among actual heat numbers underthe "Heat Number" * *
column. The associated chemistry-and material tesf data was not specific to the Unit 2 electroslag weld * *
We have* since discovered a GE transmittal letter informing
. CE Co that the electroslag data was limited to the weld procedure qualification dat *
The ISi & Materials Group Supervisor accepted the report based on a-review of the methodology use *EVENT CHRONOLOGY (Continued)
Attachment 1 Page 4 of: 15 t 1/28/88 CE Co responded to Generic Letter 88-11, informing the *N RC that vessel specific PT curves were being calculated.. New curves
- would be submitted in a Tech Spec ame.ndment reques /89 CE Co contracted GE to calculate new* PT.curves in accordance with Reg. Guide 1.99 Rev. 2. GE, with CECo's knowledge, used the 10/88 report, along with other information for curve preparatio Based upon the 10/88,report, GE considered the Dresden 2 PT data separate from the other units and prepared Dresden 2 specific PT curves. GE did not verify the inputs used from the 10/88 report. '
10/23/89 CE Co reviewed the methodology and data inputs supporting the PT curves prepared by GE, found them to be acceptable, an submitted a Tech Spec Amendment Beques /23/90 CECo responded to an NRC request for additional beltline materials information. The 1985 GE Summary Data Tables used in the evaluation of Generic Letter 88-11 were transmitted to the NBC t be reviewed in conjunction with the new PT curve /90. An *unrelated problem was found with cracking in the Quad Cities 2 reactor vessel head cladding. This led CECo to purchase a hard copy of B&W fabrication records. for the Dresden 2 and 3 and Quad, *
Cities 1 and 2 reactor vessels. *
CE Co initiated a review of the data for the upper vessel and head regions looking for any mismatches at circumferential welds joining shell courses similarto'those found at Quad Cities Unit *
This review did not include the longitudinal seam welds of the beltline area of the v.essel nor review of any materials propertie *
09/0S/91. The I_ech Spec Amendment was granted.*
03/06/92 Generic Letter 92-01,!'Reactor Vessel Structural Integrity, Rev~ 1 1 OCFR50.54(f)" was issued which required further information on the fabrication of reactor vessels. Specifically requested was information on the material used in the vessel beltline regio CECo's evaluation of the Generic Letter 92-01, included a review of the B&W fabrication reci:>rds procured in 199 ZNLD/2458-:4
- ___.._..._ --*-------* --
\\
- * Attachment 1
_ Page 5 of 15 EVENT CHRONOLOGY (Continued)
06/92.* After several weeks <;>f review, a subtlety in how GE documented
- data in their 1985 summary report was identified. Based on this, CECo recognized that some of the GE.summary data had been
.
misunderstood and that the data was not specific to the 'Dresden 2.
. electroslag welds. _
-
.. CE Co determined that the input data used in 1989 for the Dresden
_ 2 PT curve was deficient. Specifically,*chemistry and material test.
data, and initial RTNDT for the* Dresden 2 electroslag welds were not specific values, but were from a B&W general weld procec;ture qualificatio *
,
06/26/92 In consultation with GE, it was determined that the PT curves for Dresden 2 should have been the same as the Dresden 3*curve Dresden Station was notified. *
Dresden reviewed past operating and pressure testing history of-
. Unit 2 to determine if the plant had exceeded the Unit 3 PT curve limits,.. Once, during the 12/12/~0 hydrotest, Dresden 2 ~as in a..
cond1t1on that exceeded the Unit 3 PT curve. An evaluation of this-test, performed by GE, Identified that there was ample margin against brittle fracture during the tes An ENS notification was made at 4:50 p.m. on 6/2 Dresden Station immediately* applied administrativ~ controls to ensure the Dresden 3 PT curves were used for operation of Unit *
The-Daily Operating Orders '.\\'ere revised to include use of the*
Dresden 3 PT curve *
Temporary procedure changes were issued for all applicable procedures and subsequent permanent revisions were made.
. 09/14/92 The Unit-2 Tech Spec Amendment Bequest was submitted for the revised PT curve, *
01/18/93 Reviews of the GE Summary Data Tables for electroslag welds identified other discrepancies that had occurred in GE's transfer of information from the B&W fabrication records to the Summary Tables. These discrepancies had no impact on the PT curves for any of the unit ZNLD/2458-5
-
Corrections to the Summary Tables for identified discrepancies*
have been made. * CECo is currently purchasing additional B&W fabrication records applicable to electroslag welds. Any new information identified will be incorporated into the Summary Table *
EVENT CAUSES
- Attachment I Page 6 of 15 The information in the* 1985 GE Summ.ary Data:Tables forelectr.oslag welds was unclear in that weld procedure qualification record numbers~were listed *
under the "Heat Number" column. *Also, typical chemistry and other material test data was listed as if it were representative of the Dresden 2 wel *
2.. During CECo's review of the GE Summary Data Tables, an error was made interpreting the Information documented for e!ectroslag welds. The
reviewer's background (EPRI loanee) led him.to believe thatthe numbers listed on the data sheet were actual values for the weld material used In the Dresden 2 vess~I (as it was for the other data in the Summary Tables). In reality, the numbers were from a B&W general weld procedure
- qualification. Usirig this data, the reviewer was led to an erroneous conclusion that the Dresden 2 vessel PT curves could be calculated
- dlff erently than the other vessel * The CE Co supervisory review of the 10/88 Generic Letter 88"'. 1 1 report was not as detailed as it could have been. The CE Co Supervisor relied on the reputation of the expert on loan to CECo and did not critically review his work produc.
.
- * *
. During development of the PT curves in 1989, GE did not verify the design *
inputs from CECo's 10/88 Generic Letter 88-11 report..
- Engineering and Construction procedures in place at the time (pre-1990) did
. not require an engineering consultant (in this case GE) to specify assumptions and engineering judgments used in design evaluations. Also,
- those procedures did not require detailed verification by CE Co of
- assumptions and engineering judgments used In consultant design evaluation *
ZNLD/2458-6
SAFETY SIGNIFICANCE Attachment *1 Page 7 of 15 During the months that the incorrect PT curves were.in effect, two boJt ups and two hydrostatic tests were per1ormed; These tests were reviewed usi;.g the Unit 3 PT curve as a reference. During one hydrostatic test, a non-conservative, 7. * *.
degree temperature deviation from the Unit 3 PT curve existed..
At the upper limit of the Dresden 3 PT cuive for the hydrostatic test conditions, the margin against crack initiation is greater than 3. To determine the margin against crack initiation for Unit 2, calculations were performed at the actual conditions of the 12/12/90 hydrostatic test and also at the same pressure on the Dresden 2 PT curve (see graph on following page).
- For the actual test conditions, a margin against crack Initiation greater than *
2.8 existed for the hydrostatic test. * *
- .
A margin against crack initiation greater than 2.4 would have existed If the.
vessel was tested at the upp.er limit of the Incorrect Dresden 2 PT curv *
The PT curves in use at Dresden 2 exceeded the minimum 12 EFPY allowable curve based on ASME Section XI Appendix E. Had the Unit 2 PT curves continued to be used, there would have been no impact on vessel integrit *
Based upon this review, there was no safety significance.
.
- *
Vessel integrity was not impacted.
. *
Adequate margin against crack initiation existed at all time ZNLD/2458-7
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0 DRESDEN & QUAD CITIES 12 EFPY ALLOWABLE -CURVE
&SEO ON ASME SEC XI APPENDIX E
'
T=129°f'
MARGIN> 1.:S ASME APP G ALLOWABLE BOLTUP 40~
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...---- DRESDEN 2 CURVE A 1160 PSIG T=190°F'
. MARGIN> T=183°f'
MARGIN> T=17:S 0 f'
MARGIN)2.4-
-
BOLTUP 100°f'
150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE rF) *
350 Attachment I *
Page B_of 15 Dresden 2 P-T Curve Sa.fety Margin *Evaluation
.
. *
- *
CORRECTIVE ACTIONS Attachment 1 Page 9 of 15
. *
Since.1989, Engineering and Construction has been improving the.
processes.for the performance, review, and approval of design ealculations done by both CECo and it's contractors. Oyer time, several processes have been developed and enhanced to ensure the quality of engineering document *
Improvements have been made in the oversight of A/Es and contractors and in the CECo review of vendor design product Additional guidance was provided to A/Es detailing CECo expectations
for the quality of the documents that they prepar.
..
A/E Guidebook Section 7:8 was issued in December 1989 and*
provides specific requirements for selecting design inputs for *
calculations and other design products.
.
A CE Co-GE Agreement "CE Co Engineering Evaluation.
. Requirements Quad Cities, Dresden and LaSalle" was initiated in Spring 1992 and was approved by CECo and GE in January 1993; The specification calls for GE engineering output documents to contain certain information including definition of design assumptions and engineering judgements, a listing of input documentation and a definition of the analytical methods and techniques used to perform the evaluation..
Additional guidance was provided to CECo Engineering personnel to clatify manage:l'lent expectations for the review of calculations and
.other design product *
ENC-OE-81, "Review of Assumptions and J.udgements for A/E Supplied Design Evaluations," was issued in May 1991. This procedure documents the expectations for the review of
- .assumptions.and engineering judgements in A/E design evaluation ZNLD/2458-8
-
- --- ---
--~
- CORRECTIVE ACTIONS (Continued)
Attachment 1 Page 10 of 15
.
.
Improvements have-be~n made in-the processes for ensuring t~ quality of. *
_ engineering documents prepared internally by CEC _
_ *
ENC-QE-69, "Personnel Qualifications for Safety Related Calculations and/or Design Work," was issued in October 1989, to establish requirements for Engineering and Construction personnel who will
. prepare or revie~ calculation *
- *
ENC-QE-51.D "Controlled Analysis Originated by Engineering and
- Construction," was revised in January 1990, to include a standardized.
~alculational methodology and format to establish a minimum quality standar *
TID-DS-03, "ENC Technical Information For Engineering Design Calculations" was issued in December 1992. This i_nstruction provided guidance for the preparation of engineering design calculations
- .
prepared by CECo, or for CECo by A/Es. The in$truction is applicable_
to the Nuclear Engineering Department and other CECo engineering departments such as Nuclear Fuel Services and the* Produ_ction
- .
Services Departmen *
ZNLD/2458-9
- ---- ------ --*-
CORRECTIVE ACTIONS (Continued)
Attachment 1 *
Page 11 of 15 Event Cause GE summary tables ~ere untlea CECo misunderstood the data, assuming it to be specific to 02 CECo did not perform a critical review of the 10/88 repor GE.did not verify the design input requirements for the* 1989
- PT c~rve developmen ENC procedures in 1988 did not require an engineering consultant to specify assumptions and judgments used in design evaluation *
Also those procedures did not require CECo to review those assumptions and engineer1ng judgment,
ZNLD/2458-10 Corrective Actions GE procedure EOP. 25-6, originated in 1990,
. requires an independent techn\\tal revie _The 1993 CECo-GE Agreement requires GE engineering output to include a definition of design assumptions, engineering judgments, a listing of input documentation, and a description of the analytical method and techniques use ENC-QE-.69 an.d ENC-QE-51. D re-qui re an independent review.of in-house calculation GE procedure EOP 25-6, originated in 1990, requfres an independent technical revte The 1993 CECo-GE agre~ment *requires GE eng1neering output to include a
definition of de~1gn assumption A/E.Guidebook Sect1on 7.8 provides specific requirements for selecting design input for calculations and other design product The 1993 CECo-GE agreement requfres G engineering output to include a definiti-on of design assumptions, engineering judgments, a listing of input documentation, and a description of the analytical methods and techniques use ENC-QE-81 was implemented to ensure a review of the assumptions and judgments used by the A/ **
)
Atta.chment. 1
- Page 12 of 15 CORRECTIVE ACTIONS (Continued).
CECo is currently transitioning to a new organization.
- .
The Site Engineering and Construction Manager is responsible and
- accountable for all design support engineering, modification engineering, and construction services for the statio *
Communicates performance expectations to and enforces*
.
accountability among Site Engineering and Construction personne Works with NOD engineering support groups to arrange additional.
assistance for the station, as neede *
During transition: "Design Authority" (for system design and modification
- activities as specified in Nuclear Operations Policy NOP-OA.22) will be
. assigned to the Engineering and Construction Manager, who reports to th Chief Nuclear Operating Officer. Existing Nuclear Engineering Department (NED) and Nuclear Fuel Services (NFS) Department policies and procedures will be used to delegate and implem~nt this "Design Authority".
Signature authority for approving and authorizing design and modification*
activities will continue to reside with individuals presently qualified and designated in accordance with NED and NFS policies and procedure.
.
.
.
.
In the future, when it is identified that a need exists for outside technical assistance, the following could occur:
.
.
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- Site Engineering and Construction can contract for nee.ded reso.. urces through either the Corporate Engineering and Construction Groups or a vendo *
If Corporate Engineering and Construction is Involved, -it will be their * *
responsibility to conduct the appropriate reviews of the assumptions and engineering judgments used In the design documentation and provide those to the statio If _Site Engineering and Construction directly contracts a vendor, the Site will be responsible to conduct the appropriate reviews of the assumptions and engineering judgments used in the design *
documentatio *
Site Engineering aryd Construction is responsible for providing quality
. *designs to the Station. *
.* ZNLD/24?8-11
CONCLUSIONS.
Attachment 1 Page 13 of 15 Inaccurate information was inadvertently used to support a Dre.sdeh, Unit 2 Technical Specification Amendment Reques The event o~urred due to personnel error resulting in the misinterpretatio~
of technical data. The error was subtle and only identified by CECo following several weeks of detailed review of vessel fabrication record Upon discovery, CECo promptly began an evaluation to determine the significance and ~ffect of the error. Immediate actions were taken to ensure th.at future testing would be performed using the correct PT curve. -
Since 1989, actions have been taken to Improve the processes for the performance, review, and approval of design calculations done by both CECo and it's contractors. Several new engineering programs and*
procedures have been developeq and enhanced to ensure the quality of engineering document.
- *
,
- ,.
Reactor :vessel integrity was _not impacted by this event. Total margin against crack initiation exceeded the minimum ASME allowabl *zNLD/2458-12
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