IR 05000324/1990004
| ML20033E169 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/26/1990 |
| From: | Blake J, Chou R, Economos N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20033E168 | List: |
| References | |
| 50-324-90-04, 50-324-90-4, 50-325-90-04, 50-325-90-4, IEB-79-02, IEB-79-2, NUDOCS 9003090211 | |
| Download: ML20033E169 (10) | |
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UNITE 3 STATES --
M NUCLEAR REGULATORY COMMISSION -
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REGION 11 -
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101 MARIETTA STREET.N.W.
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ATLANTA, GEORGI A 30323
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i Report Nos.: 50-325/90-04 and 50-324/90-04 i-Licensee:
Carolina Power and Light Company
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P. O. Box 1551
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Raleigh,.NC 27602
' Docket Nos.:
50-325 and 50-324 License Nos.:
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Facility Name:
Brunswick 1 and 2 i
Inspectio uc ed.
anuary 29 - February 2,1990
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Date Signed
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e Approved by:
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R o-J. J. B ake, Chief Da'te Signed M te ials and Processes Section gi neering Branch ivision of Reactor Safety SUMMARY
. Scope:
This. routine, announced inspection was performed for the purpose of reviewing the licensee's corrective action (s) on previously identified enforcement matters including inspector follow-up and unresolved items in the areas of
recirculation pipe replacement and IE-Bulletins.
Results:
By document review and through discussions / interviews with cognizant technical
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personnel, the inspectors ascertained that the licensee had taken sufficient corrective action. to warrant the closing of one violation, NOV 324/89-35-02:
three unresolved -items and seven inspector follow-up items (IFIS).
Within
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these areas, the licensee's cognizant technical personnel and site management exhibited good technical competence and responded positively to technical issues dealing with nuclear safety.
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REPORT DETAILS
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Persons Contacted
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Licensee Employees
- C, F. Blackmon, Manager, Operations
- S. H. Callis, Site Licensing Representative i
- D. J. Dorman, QA/QC Manager
- J. L. Harness, General Manager, BNP
- R. E.-Helme, Manager, Technical Support t
- J. R. Holder, Manager, Outages and. Modifications
- T. Jones, Regulatory Compliance Specialist
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- J. A. McKee, Manager, Quality Assurance i
- D. E. Moore, Unit Manager Engineering
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T. Pitchford, Lead Engineer, Recirculation Pipe Replacement Project -
- R. M. Poulk,. Supervisor, Regulatory Compliance
- R. B. Starkey, Manager, BNP
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- R. L. Warden, Manager, Maintenance Other licensee employees contacted during this inspection included engineers, technicians, and office personnel.
NRC Resident Inspector
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- W. Ruland, Senior Resident Inspector
- Attended exit interview
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Follow-up on Previously Identified Enforcement Matters and Open Items (92701)
(Closed)-URI 325,324/89-35-03, Apparent NUREG-0313, Rev. 2, Programmatic Inadequacies
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This item was. identified when the inspectors.noted that certain dissimilar i
metal welds and other inconel metal welds had not been included in the NUREG-0313, Rev. 2, intergranular stress corrosion cracking- (IGSCC),
program even though the base materials appeared to be IGSCC susceptible.
The licensee's response / corrective action stated that evaluation of these l
welds was in progress.
In addition, the licensee -stated that, pending completion of this evaluation, certain interim measures have been taken to ensure that subject welds will be examined using techniques recognized by NUREG-0313, i.e., refracted longitudinal wave examination.
Also, the inspector noted that the licensee's position on this issue was that NUREG-0313 did not specifically address the inclusion of these type welds into the inspection program. However, the licensee has elected to include the identified welds in the subject program and thereby assure the frequency and examination techniques used on these welds are acceptable.
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- In conclusion, the licensee stated that following their evaluation, a decision will be made whether to include or exclude them from the subject program.
(Closed) IFI 325,324/89-35-04, Refracted Longitudinal Wave Examinations of Inconel Welds
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This item was _ identified-when the inspector reviewed selected segments of-the licensee's ISI program and determined that certain dissimilar metal welds listed under Section XI's B-J examination category had been examined d
with ultrasonic shear wave under-the subject ISI program. Other welds in this category had been excluded on the basis of a code requirement which allows for a representative sample of approximately 25 percent of the pipe weld population in this category to be examined during a ten-year
interval.
All the subject-welds had_ been examined with ultrasonic shear
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wave during preservice inspection as required by Code.
The inspector's
concern was that, in recent years, it has been shown that ultrasonic examination of inconel similar/ dissimilar welds using a shear wave -
technique-does not always produce reliable results and,. therefore, it is
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most difficult at this point to assure weld integrity.
In an effort to
resolve this concern, the licensee examined the two identified welds in d
Unit 2 with refracted longitudinal ultrasonic wave and found them to be satisfactory.
The welds in Unit 1_ will be examined in a similar manner-
during the upcoming refueling outage #7 later this year.
(Closed)IFI 324/89-35-05, Root Radiography of Pipe Replacement Welds This item was identified during a review of nondestructive examination requirements for the recirculation pipe replacement.
The' inspector's-concern was that in-process radiography for the root of safe-end to nozzle and riser to safe-end welds, for information purposes, was left to the J
discretion of GE's welding supervisors and, as such, there was no assurance that it would be done.
To forego such an examination raises the i
possibility.that completed welds would be subject to the effects of
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residual stresses from. through wall root repairs, weld repair-related problems, unnecessary radiation exposure, and project completion delays.
By memorandum dated October 4,1989, CP&L instructed GE to revise work j
plans and the-applicable procedure to require root radiography.
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.(Closed)IFI 324/89-35-06, Completion of Stress Calculation Reconciliation l
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for Recirculation Risers L
The recirculation pipe and safe-end replacement was completed late in 1989.
The new materials are Type 316 low carbon, Nuclear Grade (NG), and stainless steel (SS) for discharge risers and safe-ends.
GE Specification i
L (or Stress Calculation) No. 23A5485, Design Report - Recirculation Pipe and Equipment Loads, Rev. O, dated October 1,1985, was for the existing i
recirculation system from and to the reactor vessel.
This GE stress
- calculation included mainfolds (header), risers, safe-ends, and other piping, and equipments.
The recirculation risers and safe-ends are part of this system.
Due to the properties of the new materials, the stress
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reanalysis-is required for this portion or the whole system.
The' inspectors reviewed Report No. SIR-89-049, " Reconciliation Report for j
Brunswick Unit 2 Reactor Recirculation System Piping Replacement,"-
J performed by Structural. Integrity Associates, San Jose, California, dated October 1989. This report contains (1) introduction; (2) design criteria; (3) component description; (4) loads and load combinations; (5) material
.i properties; (6) reconciliation analyses; (7) reconciliation evaluation; (8) summary; and (9) references. ' Super Sap computer program was used for analysis.
The loads included dead weight, pressure, temperature, and seismic. The existing and new pipe material allowable stresses are 15,900
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-psi and"17,648 psi. The analysis included the following models:
Model l'- As-built (existing) Riser / Header Assembly
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Model 2 - Replacement Riser / Header Assembly
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Model 3 - As-built System with As-Built Boundary Conditions
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Model 4 - As-built System with Replacement Boundary Conditions and
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Revised Insulation It was found that for certain loadings, stress will increase in the
replacement structures.
But the final stress ratios were all
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conservatively calculated to be less than 0.8.
Therefore, the report-l concluded that the. stresses from this analysis are in compliance with the i
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stress criteria contained in ANSI B31.1-1973 and the proposed replacement activities are acceptable.
This' item is considered closed based on the.
review'of the above stress calculation.
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(Closed) IFI 324/89-35-07, Safe-end Material Substitution from SA-182 Forgings to SA-376 Seamless Pipe This item was identified when the inspector ascertained that-the material used to manufacture the replacement safe-ends had been substituted _from-SA-182 forgings to SA-376 seamless pipe material. As such,.the inspector requested that the licensee conduct an engineering evaluation to verify
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the suitability of the substitute material for the-application and to take appropriate steps to revise the FSAR to show more accurately the material used on the replacement safe-end.
The inspector reviewed Engineering.
J Evaluation Report (EER) 89-0305, Rev. O, and the attached Nuclear Safety
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Evaluation (NSE), used to document the actions taken and those which are planned to address the aforementioned concerns.
The change to the FSAR Section 5.2.31 and related Table 5.3.1-1 is' in process due to the -
implementation of PM 89-038.
The licensee's evaluation as documented in the NSE states that the as received safe-ends are equal to those originally ordered in strength requirements and IGSCC resistance.
Therefore, it was concluded that the long term reliability and structural integrity of the reactor cooling recirculating system would not be compromised by the installation of the as-received safe-ends.
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i-i (Closed)- IFI 325/89-15-01, Pipe Support Calculation Problem for Mark
i No. 1E51-41A53 The previous inspector identified civil structural beams in common use to support Pipe Support Mark Number IE51-41A53 and three other pipe supports
in the field.
The support calculation for Mark Number IE51-41A53 included
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the civil structural beams in model and analysis to check the member.
stresses and deflections.. The loads from the other three supports were not
included in the analysis.
The loads from the additional attachments could affect the analysis and the deflections of the civil structural beams and
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support number IE51-41A53, but were not considered in the STRUDL analysis.
.l The inspectors discussed' this matter with the licensee engineer and reviewed the revised support calculation provided.
Calculation I.D. 89-053-08, Rev. F1, for Plant Modification PM 89-053 was reviewed.
- The revised calculation had added three additional loads from other
attachments to the STRUDL model and rerun it.
The member stresses and deflections for Mark No.1E51-41A53 were increased, but were. still within the allowable. limits.
Therefore, this support is qualified to have three other attachments to civil structural beams in common use.
This item is considered closed.
(Closed) URI 324/89-01-03,10 CFR 21 Evaluation of Standby Liquid Control j
(SLC), Pump Spring Failures j
This item was identified to assure that the inspector had an opportunity to review the licensee's metallurgical investigation report on the subject spring failure and the licensee's subsequent evaluation to determine 10 CFR Part 21 applicability.
Following ccmpletion of the NRC inspectionj on January 27, 1989, the licensee determined that the problem involved casting flows / surface irregularities in the pumps' suction and discharge
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valves. _ The problem areas included the valve spring " cups" and on the cylinder head extension valve stop surfaces.
The report stated that the-casting flaws caused-the valve springs to lock-up/ bend on rotation, due to water flow.
The licensee's report stated that the torsional load applied to these springs, when they lock up, can result in their failure.
Failure of these valve springs can block pump suction g discharge valve closure, which results in lower than design pump flow rate.
Also, the report stated that a pump failure would result in a failure to comply with 10 CFR 50.62(4) and could result in the inability to shut down the reactor during a ATWS.
A synopsis of this information was communicated via telephone to Region II on April 28,-1989, thus satisfying 10 CFR Part 21 reporting
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requirements.
The licensee followed up the initial notification by
. memorandum dated May 1,1989, from J. L. Harness to S. D. Ebneter.
The memorandum by enclosure described the defect and indicated that the finding was reportable per 10 CFR Part 21.
The pumps in question were described as Model TD-60 triplex, positive displacement pumps,
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manufactured by Union Pump Company of Battle Creek, Michigan.
The licensee's recommendations to prevent recurrence included:
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Inspect valves in Unit 2 pumps and take appropriate corrective action as require.,
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Inspect each stop valve to assure that suspect surfaces are defect
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Replace springs of inspected valves with six-coil springs.
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= Inspect pump piston surfaces for evidence of pitting corrosion. This condition.has been associated with pump cavitation. Take corrective action as required.
(Closed)IFI 325.324/89-01-05 Storage and Protection of Components-j t
This. item was opened when the Region evaluated and agreed with the
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licensee's denial of violation 325,324/89-01-02, as it lacked specific
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examples of nonconformances in. the report.
The acknowledgement letter
stated that-storage, conditions and the effectiveness of monitoring by the-l licensee and the site QC/QA control would be revisited in a future t
inspection, In an effort to assess existing storage conditions, the inspector toured warehouses B and H used to store safety-related components which include electric motors, snubbers, valves, threaded stock, piping fittings, bar stock and welding consumables.
The inspector noted the aforementioned
components were segregated, tagged, valves and fittings were either capped or taped 'and electric motors were stored under temperature controlled conditions.
The inspector determined that storage conditions for the subject items met or exceeded applicable standard ANSI N45.2.2 requirements.
(Closed)NOV 325,324/89-35-02, Failure to Assure that Approved Procedures Affecting Quality Contain Applicable Code, Regulatory and Other Qualitative Acceptance Criteria-l The-licensee's letter of response dated January 5,1990, has been reviewed and determined acceptable by the Region II staff.
The-inspector held discussions with the cognizant Welding Engineer and examined the corrective actions as stated in the letters of response.
The inspector-l concluded that the licensee had determined the full extent of the subject noncompliance, performed the necessary follow-up actions to correct the
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present conditions and developed the necessary corrective actions to-l preclude conditions of this problem. The corrective actions identified in the letters of response have been implemented.
(Closed) NOV 325/89-01-01 Failure to Take Adequate and Timely Corrective Action This violation was issued when the maintenance inspection team determined the licensee had not taken appropriate and timely corrective action on conditions adverse to quality of components / systems important to nuclear
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safety. The subject violation cited three examples where these conditions existed:
(1) failure to perform resistance testing and maintenance of
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motorstatortemperature;(2)nocorrectiveactiontakentocorrectanoil-leak in the upper oil reservoir of RHR motor 2C; and (3) conditions responsible' for_ significant water contamination of the lube oil in the
' HPCI turbine were not investigated and corrected in a timely manner.
i The licensee's letters of response dated June 11. and August 21, 1989, were reviewed and determined acceptable by the Region II staff.
Corrective
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actions implemented to correct the described conditions. include issuance of special. procedure OSP-89-018 to test 1 insulation resistance and other
functional characteristics on RHR and core spray (CS) pump motors during_
Unit refueling outages.
During the current Unit 2 outage, the licensee tested and found acceptable RHR motors 2A and 28.
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Core spray motor '2A was also found acceptable.
Motors requiring refurbishment included RHR-2C and CS-2B. Motor RHR-2D was replaced with a u
comparable unit S/N 88H384-00010 purchased from GE under purchase order
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P.O. 548609.
l Work order 89-ASDT1 was used for its installation.
Operability procedure PT-8.2.28 will be used for _ testing purposes.
In reference to item (3) above, the. licensee determined that water leaking past the disc of valve E41-F001 in the HPCI line #1/2-E41-61-10-606 caused contamination of the turbine lube oil.
The subject valve was changed in
Unit 1 and more recently in Unit 2, during the. present outage.
At the-
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time of this inspection, valve operability had not been verified although the valve had been stroke tested successfully.
The licensee stated that following startup, oil contamination will-be monitored and checked weekly per procedure E&RC-1145.
Sampling results will be evaluated and sampling.
frequency adjusted accordingly.
The licensee has also committed to system engineering improvements as H
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described in their response to a civil penalty regarding the failure of
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management and Technical Support personnel to identify the root cause of failures and to take prompt action (s) to preclude their recurrence..In q
conclusion, Region II has further inspections planned in this area through-
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Integrated Action Plan follow-up-for item B1 (see IFI 89-34-12).
(Closed) URI 324/88-13-01, As-Built Drawing Discrepancies for Torus External Piping Systems This matter concerned the discrepancies between as-installed and as-built conditions for two pipe supports of Torus External Piping Systems.
Support No. 2821-20VH255 had a spring can end attachment connected to a composited beam and the drawing showed it to be connected to a wide flange beam 6WF-15.5.
Support No. 2B21-58VH282 had fillet weld on two sides of
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base plate for connecting to other restraint and the drawing required fillet weld all around for this connection. The inspectors held discussions with the licensee's engineer and reviewed-the revised pipe support calculations provided.
Support Calculation No. PS-B21-125, Rev.1, for Support
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No. 2B21-20VH255; and Support Calculation No. PS-B21-187, Rev. I for Support No. 2821-58VH282 were reviewed and considered acceptable which reflected as-built conditions.
The licensee's engineer have marked the D
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.above drawings to be revised and packed with other pipe support drawings
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for issuing new revisions around March 1990. Therefore, the licensee has performed eNuote steps to solve this unresolved item and the q
. inspectors etasJdered this item to be closed.
(Closed) URI 325,324/88-22-01, Determine Adequacy of QC Inspection Procedures for Inspection of Plant Modification and Determine Adequacy of
Quality Records Documenting Inspection of Modifications This matter concerned the QC inspection procedures and record maintenances for plant modifications not meeting American National Standard Institute-(ANSI) N45.2.5-1974 which is contained in Final Safety Analysis Report (FSAR) Section 1.8.
The plant modification number PM 84-384 and 84-385 for masonry wall modifications contained a final inspection record.which was as-built drawing stamped, initialed, and dated by a QC inspector. The
inspectors discussed this matter with the licensee QA Manager and reviewed the licensee's - internal memorandum (with attachments) serial number BNP/QA/QC-88-268 of file number 2090, dated September 7, 1988. Structural
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Welding at Brunswick Steam Electric Plant (BSEP) is governed by American i
Welding Society (AWS) D1.1-83 requirements and American Society of Mechanical Engineers (ASMS)Section III, Subsection NF, for certain.few
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welds. AWS D1.1 does not require that detailed records be ke welder ID, filler metal traceability or inspection details (pt concerning
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i.e., fitup, inprocess, final).
Per Work Procedure WP-502, Welding Material Requisition and Issue Ticket (WMRIT) is not a QA record and is retained
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for-one year.
WMRIT contains the welders identification and
. certification, welding procedure. used, and weld filler material and is presented to QC inspectors for verification during inspection.
Per Work Procedure WP-115, a QC inspection can'be performed as a final inspection or combination of partial inspections and. final inspection.
A final inspection only is a single _ inspection which inspects. all items at one time and QC inspector stamps, initials and dates on the as-built-drawings.
Those stamped as-built drawings are the only records of.QC inspections in the Plant Modification Packages. A combination of partial inspections and
final inspection means that-QC inspection can be divided in to several steps
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depending on the requirements or complexity of structures, systems, or supports.
A partial inspection is performed to inspect a fraction of the i
work.
A QC inspector will mark or circle the area he inspects and stamps, initials, and dates on the drawings he uses.
A fitup inspection is always a partial inspection.
During a final inspection, the QC inspector will inspect the remaining area and review or verify the previous
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partially inspected drawings or records and stamps, initials, and dates on the. final drawings.
Those final drawings are the only records of QC inspections.
All the previous partial inspection records or drawings are destroyed after the final inspection drawings stamped. The licensee's QC inspectors perform the inspections based on the documented procedures.
Therefore, this unresolved item is considered closed.
(Closed)IFI 325,324/88-36-02, Final Summary Report for IEB 79-02
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The licensee has completed all activities and modifications for IES 79-02.
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'The licensee response to IE Bulletin 79-02 of File No. NG-3513(b), Serial No. GD-79-1739, dated July 12, 1979, was received and reviewed by NRC.
This 120-day report responded to the four requirements of IE j
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Response to the first requirement stated that initial calculations for base plate design did not account for base plate flexibility and calculations were being reviewed to assure that the effects of plate flexibility would be accounted for in the determination i
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of anchor bolt loads.
Response to the second requirement stated that the design load of anchor bolts was being reviewed to assure that the load is J
1ess than the maximum allowable design load and the factor of safety (five l
for the shell-type anchors tnd four for wedge-type anchors) is in j
i compliance with the Bulletin.
Response to the third requirement stated that all anchors would be pretensioned to a load equal to or greater than the maximum allowable design lo6d to satisfy cyclic requirements.
Response to the fourth requirement stated that a testing program had been initiated per Appendix A Paragraph a, of the Bulletin due to insufficient i
documentation.
Tnerefore, the licensee's previous activities, analyses,
and calculations for the anchor bolts and base plates did not meet the.
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Bulletin requirements.
But the licensee committed that all the anchor j
bolts and base plates will meet the Bulletin requirements through:
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the review, evaluation, analysis, and revision for the calculations; (2)
testing program; and (3) modifications.
The licensee is requested to submit a final summary report to state its methods and actions taken to meet the Bulletin requirements and the licensee's commitments.
The
. inspectors and the licensee's responsible engineer had reached an agreement in the formation of-final summary report.
The licensee agreed to submit this final summary report for IEB 79-02 by end of February 1990.
The inspector will review the report when it is received. Therefore, this item is considered closed based on the agreement.
(Closed) IFI 325,324/88-36-03 Hilti Anchor Bolts Allowable Review and Justification per IEB 79-02, NRC Information Notices Nabers 86-94 anri 88-25 This matter concerned the Reduction in Hilti Anchor Bolt Allowable 1.oads from the manufacturer's catalog capacity as described in NRC information notices.
helti anchor bolts were used in both units.
IEB 79-02 requires the licensee to determine the anchor bolt allowables based on the testing capacity in the field or catalog capacity published by the manufacturer if QA documentation is available.
NRC Information Notices 86-94 and 88 25 request the licensees who use Hilti anchor bolts to review the allowables used in design calculations since the NRC has found that the test capacity in the field was 30% to 40% below catalog capacity due to variations in concrete mixtures.
The inspector requested that the lice w m review the
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Hilti anchor bolt allowables per IEB 79-02 and NRC Informanesi Notice Numbers 86-94 and 88-25.
The licensee's engineer is currently working on this item, and the licensee will submit the results of the evaluations to i
NRC for review by the end of April or May 1990. Based on the licensee's action being taken, this item is considered closed.
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Exit Interview I
The inspection scope and results were sumarized on February 2,1990, with those persons indicated in paragraph 1.
. The inspectors described the
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areas-inspected and discussed in detail the inspection results listed below.
- Proprietary infonnation is not contained in this report.
Dissenting coments were not received from the licensee, f
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