IR 05000315/1989022
| ML17328A133 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/30/1989 |
| From: | Danielson D, James Gavula, Jeffrey Jacobson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17328A131 | List: |
| References | |
| 50-315-89-22, 50-316-89-22, NUDOCS 8909120272 | |
| Download: ML17328A133 (17) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/89022(DRS);
50-316/89022(DRS)
Docket Nos.
50-315; 50-316 Licensee:
Indiana Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 Licenses No.
D.C.
Cook Nuclear Power Station, Units 1 and
Inspection At:
D.C.
Cook Site, Bridgman, MI 49106 American Electric Power Service Corporation, Columbus, Ohio Inspection Conducted:
July 10-12, 21, and August 1-3, 1989, at Bridgman, Michigan August 14 and 15, 1989, at Columbus, Ohio Inspectors:
ohn M.
son WMn James A. Gavula g
3a gp ate Date Approved By:
DE H. Danielson, Chief Materials and Processes Section Date Ins ection Summar Ins ection on Jul 10-12
and Au ust 1-3 and 14-15 1989 (Re orts No.
50-315 89022 DRS
. 50-316 89022 DRS p
i f
snspectson efforts on Units 1 and 2 (73052, 73755)
and corrective actions to NRC violations concerning design control (92702).
Results:
Two violations were identified; failure to follow procedures-Paragraph 6 and inadequate design control measures for checking and verifying the design - Paragraph 6.
Based on the results of the inspection, the following was noted:
The inservice inspection effort was carried out in.a professional manner and documentation was judged to be excellent.
8909120272 S90905 PDR ADUCf; 05000315
The types of design control deficiencies identified during this =inspection are similar to deficiencies noted in the previous NRC Engineering Team Inspection.
Even though the corrective action programs are in early stages of implementation and the deficiencies were not safety significant in nature, the proper perspective on "attention to details" needs to be emphasize DETAILS 1.
Persons Contacted American Electric Power (AEP)
+D. Williams, Jr., Senior Executive Vice President, Engineering and Construction
+M. Alexich, Vice President, Nuclear Operation
+"W. Smith, Plant Manager
"J.
Sampson, Safety and Assessment Superintendent
"R.
Rickman', ISI Supervisor C. Freer, ISI Coordinator J. Fitchuk, NOE Technician
+S.
Brewer, Nuclear Safety and Licensing Manager
+"B.
P.
Lauzau, NS8L Engineer
+"A. Dey, Design Engineer
+T. Kwiatkowski, Design Division Staff
+P. Barrett, Director, QA U.S. Nuclear Re ulator Commission (NRC)
"B. Jorgenson, Senior Resident Inspector
"J. Jacobson, Reactor Inspector, Metallurgical
"J. Gavula, Reactor Inspector, Mechanical
"Denotes those attending the exit meeting on July 21, 1989.
+Denotes those attending the final exit meeting on August 15, 1989.
2.
Licensee Action on Previous Ins ection Findin s (92702)
a.
(Cl os ed) Viplati on (315/88028-01B 316/88032-01B):
Failure to specify fs et we size or soc et we s.
s et weld size must be specified by design to assure welded joint integrity.
A review of modification packages (RFC's)
by the NRC disclosed a failure to specify fillet weld size for socket welds.
As corrective action the licensee has incorporated joint details specifying weld size for socket welds, flange welds, half couplings, and integrally reinforced fittings into the AEP Welding Manual.
These joint details are implemented by plant procedures.
The NRC inspector reviewed the weld joint details and found them acceptable.
b.
(0 en) Violation (315/88028-06.
316/88032-06:
Failure to provide adequate dna snspectson documentation to etermine that the proper fillet weld size for socket welds was installed.
The documentation did not reference the appropriate inspection procedure and furthermore, the procedure required the inspector to perform a weld size calculation which was not documented.
As corrective action, the licensee inserted the appropriate weld joint details, specifying weld size, into the inspection procedure.
The NRC inspector reviewed
this revision and determined this to be acceptable.
Additionally, a sign-off indicating the appropriate procedure utilized for the inspection will be incorporated in the documentation.
As a result of the lack of up front specification of fillet weld size (Paragraph 2. a) and the lack of adequate inspection documentation, the installed weld size cannot be determined for work previously completed.
This violation will remain open pending verification of previous work.
(Closed)
0 en Item (315/88028-05 316/88032-05
The 1967 Edition of USAS 831. 1 Power Piping requires a soc et weld fillet size of 1.25 times the pipe wall, while later editions of the code reference a fillet size of 1.09 times the pipe wall.
The licensee's inspection procedures referenced the 1.09, however, FSAR commitment is to the 1967 edition (1.25),
While it is recognized that it may not be possible in all cases to provide the specified 1.25 due to fitting limitations, the licensee must resolve this deviation.
The licensee elected to revise both the AEP Welding Hanual and appropriate inspection procedures to reference the 1.25 factor.
To avoid field installation problems, a maximum weld leg size equal to the thickness of the socket wall is specified.
The NRC inspector reviewed these revisions and found them acceptable.
(0 en) Violation (315/88028-01F 316/88032-01F:
An unverified ana ytsca techn>que was use to evaluate a structural component for several piping supports.
This problem originated when a finite element analysis program for piping was used to analyze a structural member.
In trying to model the structural member, the analyst did not properly correlate member properties and subsequent checking and design verification activities did not disclose this deficiency.
As a corrective action, the licensee procured a general purpose structural analysis program that will be used in the future.
The NRC inspector reviewed the new analysis of the specific component that utilized the structural finite element program.
In addition, several other pipe support analyses using the structural program were reviewed by the NRC inspector.
No concerns or deficiencies were identified during these reviews.
Insofar as eliminating the potential for misapplication of the piping analysis program in the future, the licensee s action appears to be adequate.
However, additional corrective actions to strengthen the design verification process across all disciplines, such that other inappropriate analytical techniques are not used, needs to be assessed.
This aspect should be addressed during the licensee's upcoming detailed audit of the design control process as committed to in AEP's June 2, 1989, response.
>>Pending an NRC review of the licensee s audit findings and subsequent corrective actions, this item will remain ope Unit 1 Inservice Ins ection (73052 73755)
Inservice inspection (ISI) of Unit 1 components was performed by Southwest Research Institute (SWRI) during the March through May, 1989, refueling outage.
This examination of selected Class 1 and Class
components constituted the second ISI of the first period of the second ten year interval of operation.
Selected components in the following areas were examined:
Reactor Pressure-Vessel
'team Generator Reactor Coolant, Piping Chemical and Volume Control Piping Reactor Coolant Pump Safety Injection Piping
.
Containment Spray Piping The NRC inspector reviewed the following SWRI NDE procedures:
Title Solvent Removable Liquid Penetrant Dry Powder Magneti c Par ticl e Manual Ultrasonic - Pressure Vessel Manual Ultrasonic - Studs and Bolts Manual Ultrasonic - Austenitic Piping Manual Ultrasonic - Longitudinal Wave Mechanized Ultrasonic - Austenitic Piping Mechanized Ultrasonic - Inside Vessel Manual Ultrasonic - Austenitic Thin Wall Piping Inner Surface Stud Access Holes Manual Ultrasonic - High Attenuation Materials Visual Examination Procedure No.
(SWRI-NDT)
200-1/70 300-1/35 600-15/72 600"18/42 600-31/24 600-49/1 700-10/6 700-11/10 800-36/40 800"104/3 800-126/0 900-7/14 All procedures were found to be in accordance with ASME Section XI, 1983 Edition, Summer 1983 Addenda.
The NRC inspector also reviewed the certification records of SWRI examination personnel and found them to be in accordance with SNT-TC-1A, 1980 Edition.
The results of the ISI were reviewed by the NRC inspector and found to be acceptable.
No rejectable indications were found during this examination.
Eddy Current inspection (ECT) of the steam generator tubing was performed by Westinghouse.
The extent of the inspection was as follows:
Steam Generator 1 - 531 tubes tested full length 2723 tubes tested from hot leg thru U-bend Steam Generator 2 - 507 tubes tested full length 2752 tubes tested from hot leg through U-bend
Steam Generator 3 - 515 tubes tested full length 2804 tubes tested from hot leg through U-bend Steam Generator 4 - 513 tubes tested full length
.
2756 tubes tested from hot leg through U-bend Some additional examinations were performed as required.
The examination was performed utilizing multifrequency; multiparameter digital data collection equipment (HIZ-18A).
The examination was performed in accordance with Procedure MRS 2.4.2 Gen-28, Revision 3,
"Digital Nultifrequency Eddy Current Inspection of Heat Exchanger Tubing."
This procedure was reviewed by the NRC inspector and found to be acceptable.
Certification records were also reviewed and found acceptable.
This examination resulted in the following tube plugging:
Steam Generator Number Number of Tubes Plu ed
2
4
40
57 The NRC inspector reviewed the test data on a sample basis and concluded that the data was evaluated in a conservative manner.
Those tubes removed from service were plugged utilizing the Westinghouse mechanical plug.
4.
Unit 2 Inservice Ins ection Inservice examination of Unit 2 components and Reactor Pressure Vessel (RPV) was performed by SWRI during the 1988 refueling/steam generator outage.
The RPV examinations were performed to complete the first ten year interval requirements.
The Class 1 and Class 2 component examination constituted the first ISI of the second ten year interval.
Selected components in the following areas were examined:
Pressurizer Reactor Coolant Piping Safety Injection Piping Chemical and Volume Control Piping and Tank Residual Heat Removal Piping and Heat Exchanger Reactor Coolant Pumps Regenerative Heat Exchanger Containment Spray Piping Feedwater Piping
'
The RPV examination included the following areas:
Circumferential Welds Longitudinal Welds Flange Welds Nozzle to Shell Welds Nozzle to Safe end Welds Lower Head Meridional Welds Nozzle Inside Radius Vessel Interior Surfaces Core Support Structures Core Barrel Upper Internals The NRC inspector reviewed the following SWRI NDE procedures in addition to those listed for the Unit 1 examination:
Title Mechanized Ultrasonic - Austenitic Piping Mechanized Ultrasonic - Inside Vessel Visual - Reactor Internals Manual Ultrasonic - Thin Wall Vessels Manual Ultrasonic - Heat Exchanger Welds Manual Ultrasonic - Pressure Vessel Welds Manual Ultrasonic - Ferritic Piping Manual Ultrasonic - High Attenuation Austenitic Materials Procedure No.
(SWRI-NDT)
700-10/5 700-11/9 900-2/12 600-26/12 600-30/21 600-33/11 600-41/15 800-17/33 All procedures were found to be in accordance with ASME,Section XI, 1983 Edition, Summer 1983 Addenda.
The certification records of SWRI examination personnel were also reviewed and found acceptable.
The results of the ISI were reviewed by the NRC inspector and found to be acceptable.
The only recordable indication, other than geometry, was detected in the nozzle-to-shell weld of inlet Nozzle 2-N2-1.
When sized in accordance with ASME Section XI sizing procedures, the indication exceeded the acceptance criteria of IWB-3512.
This examination was conducted from the inside vessel surface with a metal path of approximately 15", utilizing Code required amplitude drop.
The licensee invoked the provisions of IWB-3200 which allows for supplemental examinations to evaluate flaws.
The supplemental examination was performed from the outside surface with a metal path of approximately 2".
Utilizing several different sizing techniques, it was concluded that the indication was a
Code allowable, fabrication-induced, flat surfaced, planar slag inclusion.
The NRC inspector reviewed the examination data and found it acceptabl A preservice examination of selected Steam Generator (SG) components installed during the Steam Generator Repair Project was performed by SMRI.
The UT examinations were performed prior to and after hot functional testing of the SG's.
The NRC inspector reviewed applicable procedures and personnel certifications and found them acceptable.
The following examinations were performed:
SG Nozzle to Elbow Welds Tube Sheet to Head Welds Nozzle Inside Radius Sections Tube Sheet Stub Barrel Meld (SG No. 4)
Stub Barrel to Shell
The examinations revealed two recordable indications.
A weld inclusion in SG No. 4, tube sheet to stub barrel weld was detected.
This was determined to be a spot indication which is acceptable in accordance with Table IWB 3511-1 of Section XI.
Seven indications were detected on the Nozzle to Elbow (STM-23-03) weld of SG No.
3.
These were all determined to be Code acceptable per IMA-2232 and IWB-3514.4 of Section XI.
The NRC inspector reviewed the pre-service data and found it acceptable.
Coolant Loo Weld Mock-U As a result of extensive welding problems associated with the Unit 2 coolant loop welds to the new steam generators, (See NRC Inspection Report No. 50-316/88026)
the licensee was requested to demonstrate effective ultrasonic detection of flaws in the mid plane regions of these welds.
In an effort to demonstrate the examination technique, the licensee retained the services of SWRI to fabricate a pre-cracked mock-up of the loop welds.
A specially flawed, dissimilar metal weld mock-up per SWRI Orawing E-3378-611 was fabricated.
The cracks were simulated by a technique involving sequence welding to form cavities at pre-selected axial positions and thru-wall locations.
These cavities were then filled with a crack susceptible filler metal, then welded over.
This technique provided five ID surface cracks of varying depth and four mid plane cracks of varying depth and dimension.
The NRC inspector reviewed the SWRI ultrasonic test report of the mock-up, as well as radiograph verification, and found that all rejectable flaws were located with good resolution.
This serves to demonstrate the ability to locate Code rejectable flaws in the mid plane region of the coolant loop welds for ISI purpose ~
~
6.
Review of Re uest for Chan e
(RFC)
No. 3029 a ~
b.
B~kd During the Unit 1, ISI Component Support Inspection, multiple problem reports were generated identifying pipe supports with configurational discrepancies.
The RFC requested authorization for minor modifications to restore the deficient supports to either an as-designed condition or.to an approved alternate condition.
As documented in 'an AEP internal memo from A. K. Dey to R. Rickman dated Hay 18, 1989, out of 201 pipe supports inspected during ISI, 66 were found to have discrepancies between the as-designed and as-found condition.
Of these 66 supports, 20 required field modification in order to meet the design basis limits.
Of these 20 supports, 10 were returned to the or iginal design condition, whereas the other ten required varying degrees of redesign and modification in order to meet the design basis limits.
In addition to the ISI supports, an unspecified number of other supports were inspected during the resolution of the above ISI discrepancies.
From these inspections an additional 16 discrepancies were documented, of which two required field modification in order to meet the design basis limits.
Also, while reanalyzing and evaluating piping systems and supports in order to resolve all of the above discrepancies, two other supports were determined to have been inadequately designed and required modification.
Furthermore, one additional support had to be added to the Component Cooling Water System in order to meet the design basis limits.
In addition to all of the above, two new small bore pipe supports were added to the Diesel Generator Starting Air System to correct an undisclosed problem.
Review of Calculations The NRC inspector reviewed the following calculations for compliance with licensee commitments and NRC requirements.
These calculations were for support modifications and they were performed by site engineering since the Region III Engineering Team Inspection which was completed in March 1989.
DC-D-1-GFW-L210 (Feedwater S stem)
The original ISI observation stated,
"Vertical angle has missing welds and cracked due to pipe movement."
An operability analysis was subsequently performed assuming that the vertical angle with the cracked weld provided no horizontal restraint.
In addition, during subsequent reviews of other pipe supports on the system, it was found that Support 1-GFW-L812 (Feedwater System)
was modelled as a
two way restr aint even though it provided restraint in all three directions.
This modelling discrepancy was also corrected in the
operability analysis.
The results of the above calculation indicated that the system could meet the interim operability allowable stresses as established in the AEP correspondence from H. P. Alexich to A. B. Davis (NRC) dated March 13, 1989.
However, additional evaluations of support 1-GFW-L812 concluded that the support would significantly exceed the FSAR stress limits.
This deficiency was attributable to the original design of the support and was not necessarily related to the above noted mis-modelling of the support.
A support modification was issued to the field and the deficiency was subsequently corrected.
The NRC inspector noted two problems with the way in which the above operability evaluation was handled.
First, the discovery of the overstressed pipe support was never documented in a condition report as required by D. C.
Cook Procedure PNI-7030, Paragraph 4.5. 1.
This paragraph states,
"Problems, as defined by this procedure, or any other condition adverse to quality of nuclear safety related systems (including both equipment and activities), shall be reported using the Condition Report."
The deficient condition for the above support as well as for those related to supports 1-ACA-R911 and 1-ACA-R910 (Control Air System),
1-ACCW-R924 and 1-GCCW-L181 (Component Cooling Water System)
were not documented on Condition Reports.
Failure to accomplish activities in accordance with the above prescribed procedure is an example of a violation of 10 CFR 50, Appendix B, Criterion Y (315/89022-01; 316/89022-01).
In addition, even though the support (1-GFW-L812) was significantly overstressed, no evaluation had been performed to determine the degree of overstress.
The complicating factor in this situation is that this support is also shared by another system as well.
On this basis, the past operabi lity of both systems had not been determined.
Pending completion of this operability evaluation, this is considered an Unresolved Item (315/89022-02; 316/89022-02).
DC-D-1-GCCW-L265 (Com onent Coolin Water S stem)
The original ISI observation stated,
"Several areas lacking full thread engagement; rods are bent."
Additional inspection revealed that although the support would have an uplift load during a design basis event, it did not provide any uplift restraint.
An operabi lity analysis was subsequently performed using increased damping as prescribed by the above interim operability acceptance criteria.
With the increased damping, no uplifting was predicted to occur and the system was considered operable.
A support modification was issued to the field and the deficiency was subsequently corrected.
DC-D-01-FW-RII (Feedwater S stem)
The original ISI observation stated,
"No vertical angles, shim is 1 1/4" instead of 1/2" thick as called for on drawing."
In
addition, supports 1-AFW-R929 and 1-AFW-R930 were noted as missing the top angle such that they could not restrain the pipe in an uplift direction.
The operability analysis was adequately performed and modifications to the supports were made in the field to return them to the as-designed condition.
DC-D-1-CCW-L181 (Com onent Coolin Water S stem)
During the review of the component cooling water system analysis to address a different deficiency, it was noted that this support was modeled as a vertical restraint only, when in fact the support provides restraint in all the directions.
The support was subsequently redesigned to provide only the restraint in the vertical direction.
In doing this redesign, the analyst used a pinned arrangement for the support.
The analysis of the pin failed to evaluate the bending induced into the pin, and the design control measures for verifying or checking the adequacy of the design did not disclose this deficiency.
Failure to provide adequate design control measures for verifying and checking the design is an example of a violation of 10 CFR 50, Appendix B, Criterion III (315/89022-03A; 316/89022-03A).
A preliminary evaluation of the above configuration by the licensee concluded that there was no safety significance associated with this occurrence.
DC-D-1-CS-R541 (Chemical and Volume Control S stem)
The original ISI observation stated,
"Base plates are 90'ut of position."
The evaluation of this discrepancy concluded that the as-found support met FSAR criteria.
The review by the HRC inspector disclosed that only two of the possible four load vector combinations were evaluated by the analyst.
The two load vector combinations for the weakest direction of the support were not evaluated.
Failure to provide design control measures for verifying and checking the design is another example of a violation of
CFR 50, Appendix B, Criterion III (315/89022-03B; 316/89022-03B).
A preliminary evaluation of the above deficiency by the licensee concluded that there was no safety significance associated with this occurrence.
The above discrepancies are very similar to the types of discrepancies previously identified in NRC Inspection Reports 50-315/88028; 50-316/88032.
The corrective actions taken in response to the violations cited in the previous report include the performance of a detailed audit of the design control process.
This task is scheduled for completion in September 1989.
However, interim actions, which included "in-depth discussions with engineers/designers on the problems that have been identified and on the expectations to produce high quality work," may need to be re-emphasize.
Unresolved Items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an open item, a
deviation, or a violation.
An unresolved item disclosed during this inspection is discussed in Paragraph 6.
8, Exit Interview The Region III inspector met with the licensee representatives (denoted in Paragraph 1) on August 15, 1989, and at the conclusion of the inspection on August 15, 1989.
The inspector summarized the purpose and findings of the inspection.
The licensee representatives acknowledged this information.
The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed during the inspection.
The licensee representatives did not identify any such documents/processes as proprietary.
12