IR 05000324/1989015
| ML20247F337 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/12/1989 |
| From: | Blake J, Chou R, Economos N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20247F301 | List: |
| References | |
| 50-324-89-15, 50-325-89-15, IEB-87-002, IEB-87-2, IEB-88-005, IEB-88-5, NUDOCS 8907270128 | |
| Download: ML20247F337 (10) | |
Text
. gig Y UNITED STATES V ,)c." MCEOb-e -
.. . NUCLEAR REGULATORY COMMISSIOW
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w. i_ '[ REGloN il, 101 MARIETTA STREET, * *#- ATLANTA, GEORGI A 30323 g... . Jo b
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-Report Nos,n 50-325/89-15;a'nd 50-324/89-15
.L Licensee: Carolina Power,and Light Company
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P. 0. Box 15511 Raleigh,,NC 27602 Docket Nos.: 50-325 and 50-324 License Nos.: -DPR-71.and DPR-62
[ Facility Name: JBrunswick 1 and 2 Inspection C . ne'19-23, 1989 Inspecto s: / s 7 69- 'os Date Signed
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Approved'b / '/
J Blake, Chief Date Signed at ials.and Process Section En ineering Branch'
Division of Reactor Safety SUMMARY-Scope:
- This routine, announced. inspection was in the areas of Reactor Core Isolation
. Cooling - (RCIC); F013 Valve - Replacement; Temporary Instruction (TI), 2500/27 Fastener Testing to Determine Conformance with Applicable Material Specifica-tions; IEB 88-05, Nonconforming Material Supplied by Piping Supplies, Inc. . . . . . ; review of inspector identified items and -Licensee Event Reports (LERs).
'Results:
In the areas inspected, violations or deviations were not identified. The main
' objective. of this work effort was to evaluate the licensee's handling of .the RCIC F013 valve ' replacement and the resulting relocation of the ASME Code, Class-1 boundary. By work observation and document review, the inspectors ascertained that ; management had shown strength in allocating sufficient ;
resources .to. assure that design issues, material procurement, fabrication and
, testing associated with this modification were handled satisfactoril This s H,
management involvement / strength allowed the work to progress in an orderly manner and without significant problems, t .
L 8907270128 890713 I .PDR ADOCK 05000324 L Q PDC E
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REPORT DETAILS Persons Contacted Licensee Employees E. Betz, Level III NDE Examiner-
- J. M.. Brown, Resident Engineer (NED)
- A. G. Cheatham, Manager, E&RC B. Cruise, Outage Management Integrated Schedule Section (0MISS, Engineer)
- L. E. Dorman, Director, QA/QC
- K. E. Enzor, Regulatory and Compliance Director R. Hanford, Materials Engineer, Corporate
- J. L. Harness, General Manager R. D. Harrington, Senior Engineer, Corporate J. C. Hopkins, Stress Analyst, Corporate R. LaBaw, Design Engineer (NED)
J. McCracken, OMISS, Modification Engineer R. M. Paulk, Project Specialist, E&RC V. K. Stephenson, Project Engineer for Stress Analysis, Corporate i R. Tripp, Supervisor, E&RC '
Other licensee employees contacted during this inspection included operators and administrative personne NRC Resident Inspectors W. Levis, Resident Inspector
- Ruland, Senior Resident Inspector
- Attended exit interview Licensee Action on Previous Enforcement Matters IE Bulletins (IEB) Units 1 and 2 (92703)
'(Closed) IE Bulletin 88-05, as modified by Supplements 1 and 2
' Nonconforming material supplied by Piping Supplies, Inc. at Folsom, New Jersey, and West Jersey Manufacturing Company at Williamstown, New Jerse By memorandum, serial number NLS-88-219, dated September 7,1988, the licensee responded to NRC Bulletin 88-05, as modified by Supplements 1 and The licensee's records search showed that no flanges or fittings were purchased from Piping Supplies, Inc. (PSI). i However, the licensee found that he had purchased 148 pipe flanges which were manufactured by West Jersey Manufacturing Company (WJM),
from SA-105 carbon steel material. These flanges were purchased for
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~2 use in safety-related system Of the 148 flanges, the licensee .
determined the following: i Flanges Scrapped 13 Flanges in Stock 16 Flanges Removed by Plant Modifications 12 Accessible Flanges 14
. Inaccessible Flanges J Total Number of Flanges Located as of August 3, 1988 Suspension 62 Total Number of Flanges Not Yet Located as of August 3, 1988 Suspension 86 Total Number of Flanges Identified as of August 3, 1988 Suspension TM All of the 14 installed accessible flanges were tested for hardness using an Equotip Portable Hardness Tester. Fifteen of the 16 flanges
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in stock were also tested for hardness in the same manner, prior to suspension of activities per Supplement 2. These flanges were all found to be within the acceptable hardness rang Flanges in stock have been placed on hold (per Supplement 1 directives). Detailed information concerning material specification, component, nature, size, pressure rating, and vendor as required by the Bulletin and' supplements has been documented and appears to be in orde In reference to this matter, the inspectors reviewed a report, l entitled NUMARC 88-01, issued on the subject IEB by the Nuclear Management and Resources Council (NUMARC). The report was attached to a cover letter addressed to Mr. F. J. Miraglia of the NRC from W. H. Rasin, Director, Technical Division, dated October 27, 198 The report summarized the results of field testing, laboratory testing, and stress analyses performed in response to the subject IE Bulletin. The report concluded that the observed nature and degree of nonconformance was limited, and did not represent a safety concern, i.e. , over 99.5 percent of components tested exhibited test results indicative of the correct ASME/ ASTM specified materials. The ,
program identified 25 carbon steel blind flanges which did not meet i minimum tensile strength requirements, i.e., 45 ksi as opposed to the 70 ksi required by code. However, stress analysis calculations show that for blind flanges, tensile strengths in the order of 45 ksi are sufficient to accommodate stress allowable with margin. Based on these findings, NUMARC recommended closure of this issue without further utility action requirement I r
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(Closed) IE Bulletin 87-02/ Temporary Instruction 2500/27 i
Inspection Requirements for NRC Compliance Bulletin 87-02 " Fastener Testing to' Determine .Conformance with Applicable Material Specifications" i
In response to.the subject Temporary Instruction (TI), the inspectors discussed corrective actions / disposition applicable to fastener sample BS07. Stu, identified as ASTM /SA 193 GRB8 stud materia The licensee's~ investigation had determined that these studs were made from material which contained higher amounts of carbon and sulfur than that specified by the standard. The fasteners were purchased in 1985 as a spare lot of eight set screws for the RHR pump. The .(fasteners) had been purchased from Byron Jackson who had, in turn, ,
bought them from Texas Bolt as an off-the-shelf item. The licensee indicated the fasteners were received with ' a certificate-of-compliance; hence, it was not possible to verify chemistry and/or mechanical properties by quality, record review during receipt 1 inspection All _ eight -set screws have been discarded and replaced j with others meeting the applicable specificatio . Licensee Event Reports (LERs) (92701)
(Closed) LER 88-30, Rev.1, Unit 1 and LER 88-05, Rev.1, Unit 2, Safety Relief Valves' Setpoints Exceeded During Testing at Wyle Laboratories This event report was initially communicated to Region II on November 30, 1988, and subsequently revised on March 1,1989. The licensee's report ,
indicated that during the Unit 1 1988/1989 refuel / maintenance outage, !
testing of safety relief valves (SRVs), in accordance with Technical Specification (TS), revealed that 9 of the 11 valves tested lifted at pressures outside the TS tolerance of +/- 1%. Each of the nine valves lifted at pressures above their setpoint. Four of the nine (821-F013B, C, H, and L) setpoint drifts are attributed to corrosion induced bcnding of the pilot valve disc to seat surfac The other five (G21-F013A, E, G, J, and K) were attributed to labyrinth seal frictio The' tests on the subject SRVs were performed by Wyle Laboratories. Both Brunswick Units use Target Rock, Model No. 7567F, two-stage SRVs. To remedy this valve disc corrosion problem and improve SRV service life, the licensee contracted Target Rock to rebuild the SRVs and replace the existing pilot valve discs with new ones made of precipitation hardening stainless steel PH13-8Mo material. Discussions with cognizant licensee personnel disclosed that five SRVs were refurbished with discs made of the i new material. The SRVs were decertified and reinstalled in Unit 1 during the 1988/1989 refueling / maintenance. A similar maintenance work was performed on six SRVs in Unit 2. Eventually, all SRVs in both Units will l be refurbished with the aforementioned replacement materia The inspectors reviewed testing and decertification records for content and accurac i
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4 RCIC Valve F013' ' Replacement and Relocation of the ASME Code, Class 1
_ Boundary on the Related Piping. Unit'1 (37700) (55050) (57090)
The. . inability. to establish valve closure, that satisfied containment isolation requirements, with the motor operator on valve 1-E51-F013, led the licensee to take immediate corrective. action. .By document ~ review and discussions with. cognizant engineers, the inspectors. ascertained that corrective action ' was implemented: by modification no.89-053 which required the installation of- a new, identical gate valve immediately'
upstream of the existing valve with the existing valve to be left in place and manually secured / locked in the open~ positio To accommodate : .l installation of the new valve and to satisfy ASME Code Class 1 boundary . !
requirements, replacement piping required for .the modification was upgraded accordingly. A new 3/4" drain / test connection with double isolation valves was installed in between the old and new valve to fulfill Class 1 requirements. Details relative to design, loads, material and testing considerations are as follows:
Valve 1-E51-F013 is located on the discharge side of the RCIC pump upstream -of the connection into the feedwater line. This valve functions as a .normally closed containment isolation valve and it is automatically opened.on reactor vessel low water level indication if Main Steam' and Feedwater are both isolated to ensure inflow to the reactor vessel to cool and cover the core. The subject valve.is-required to open against a differential pressure of 1,140 psid and close against a differential of 440 psid. Design pressure and temperature for this valve is 1,500 psig and 425 F, respectivel Valve opening and closing time is required to be within 20 second To satisfy these design' requirements, valve F013 is a 4," Class 900, Flex Wedge, motor operated gate valve. The replacement valve was manufactured by Anchor / Darling Valve Company and procured by CP&L under purchase order 435013 for ' Shearon Harris. The valve is identified by Serial Number E9074-67-3. In that this valve was originally purchased to ASME Code Class II requirements, the licensee contracted Anchor / Darling to perform an engineering evaluation and determine its suitability for a Class 1 applicatio Anchor / Darling Report No. R89.064 dated June 1,1989, certifies the subject valve meets ASME Code (71S72) Class 1 requirements. The new 3/4" .
drain / test connection isolation valves are ANSI Class 1500#, Globe drain valves manufactured by Convail In Licensee generated documents relative to this modification that were reviewed by the inspectors were as follows:
Modification 89-053: RCIC F013 Valve Replacement Engineering Evaluation Report: EER No. 89-0178, Rev. 0 Engineering Evaluation Report: EER No. 89-0183, Rev. 0
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The applicable codes for this modification were as follows: ,
- Materials and fabrication - ASME Code Section III, Division 1,1986, j No Addenda Welding - ASME Code Section IX, 1986 Edition, No Addenda, Inspection and Testing - ASME Code Section XI, 1980 Edition including Winter 1981 Addenda Within these areas, the inspectors reviewed the aforementioned modifica-tion package and the engineering evaluation reports focusing on material procurement, i.e., valves, piping, structural steel and welding consumables; field installation / fabrication records 'acluding welding
. l procedure qualifications, welder performance records, and weld joint ,
tickets; radiographs and surface examination reports. In addition, the j inspectors conducted field inspections to observe completed welds, obtain as-built measurements to verify dimensions /information on as-built !
drawings and bill of material Within these areas, the inspectors observed that one of the radiographs exhibited a small linear indication in the center of the weld joint which was believed to be a root condition associated with the welding process --
gas tungsten arc (TIG). Because the density of the indication was substantially higher than the surrounding region, the inspector stated that additional information was needed on the specific area of interest, weld 251 radiographic station 4-5, to permit an objective evaluation /
interpretation of the indication and, therefore, determine what impact, if any, it would have on the integrity of the join At the time of this inspection, the licensee had no means available that could be used to measure quantitatively the depth / severity of this indication. Therefore, ;
an inspector follow-up item (IFI) was identified until a more quantitative measurement could be obtained which would allow an objective evaluation of the indication in question, IFI 50-325/87-15-02, Radiographic Indication in Weld 251 Station, (4-5). .
Other areas of interest examined by the inspectors are as follows:
Because of the replacement valve and change at the Class 1 piping boundary, the licensee performed a stress reanalysis, Calculation No. 89-053-01, Rev. F0, for Isometric Drawing No. F-28046, Sheet No. 563, Rev. 3. The boundary of the stress analysis included three anchors at Node Nos. 310, 470, and 8. This analysis included approximately 55' of Class 1 piping and 40' of Class 2 piping. Thirteen pipe supports, one j which had been relocated, were included in this analysis, j The input data for computer model ADLPIPE dated May 30, 1989, were reviewed and checked against the isometric drawing, Final Safety Analysis (FSAR), USA Standard Code for Pressure Piping (831.1), and ADLPIPE User's Manua The data checked included node point designations, pipe lengths, support locations, support drawings, support types, support stiffness, stress intensification factors, valve locations, valve designations, valve
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6 H weights, valve centroid gravity locations, anchor point displacements due to pipe run, insulations and their weights, . loadings, load combinations, material allowables, pipe pressures, pipe temperatures, pipe ' material .
properties, seismic factors- for Optrating Basis : Earthquake _(OBE) .and Design Basis Earthquake (DBE). The ~ output data were also: reviewed for consistency with . input data and reliability of results. . During the input-data checking, the inspectors noted that two valve-torsional loads, due to the eccentricity of the valve center of gravity, were not considered in the computer input; and three -nodes 'which were defined in the computer; input data were not specified and located on the isometric drawing. The licensee engineer stated that the torsional ' loads, were small and tha they would not have significant-impact in the analysis. For the nodes not specified in the isometric drawing, the licensee quickly revised the-isometric drawing to show the three missing nodes at their right location The Stress Calculation No. 89-053-01, Rev. F0, was reviewed and the following information checked: references, drawings, list of' pipe
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properties, lumped . weights . at node points, valve data tables, support types and directions', design conditions and operating modes', enveloped response spectra, analysis cases, load case combinations, stress summary,
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and tables of support load summary for support design group. A major stress summary is listed below and.all stresses are acceptable:
Max I Operatin Load Stress Node Allowable Stress Condition Combination (psi) N (psi) Ratio Normal D 6477 535 16056 .403 i Normal y+ANGE R 19369 8 23368 .829 l Normal, 0+
y+NGE RA 24553 8 38340 .640 Upset P + D + OBE
+ OBE 12010 440 18000 .667 Emergency P4 0 8050BE
+ DBE 16080 440 27000 .596 s
P + D gm0BE
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Seismic + T + OBE 15829 407 30000 .528 sam Boundary 315 Nomenclature:
P = Pressure Stress Dead Weight Stress g=ANGE=ThermalStressRange R
OBE = Operating Basis Earthquake (0BE) Inertial Stress DBE = Design Basis Earthquake (DBE) Inertial Stress OBE[==OBESeismicAnchorMovementStress DBE DBE Seismic Anchor Movement Stress
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Support Calculation Review After the stress calculation was completed, the stress group sent the Support Impact Summary and Support Load Summary to a support design group to reevaluate each support for impact due to load / movement change. The Support Impact Summary requested that Support No.1E51-41FH50, a rod hanger, be relocated due to interferences with the new valve. The support design group made a load and movement comparison table for new and old loads / movements for each support in the analysis. The inspectors reviewed calculation I.D. 89-053-08, Rev. F0, dated June 19, 1989, which included new calculations and justifications for 12 supports. Review of Mark No. PS-6192 (Anchor) was not required because the anchor was designed for plastic momen The calculation included objective, assumptions, references, load summary sheets,' calculations, and conclusio Each support had an individual load summary sheet and calculation for justificatio The calculations included justification due to load and movement change, standard component allowable checks due to load increases, weld and member requalifications, etc. Most supports had load increases. But they were still within the design allowables except Anchor 1E51-41A53 which required reinforcement. The movements had slight-differences and all were acceptable. Anchor 1E51-41A53, due to a significant load increase, required STRUDL reanalysis. The computer input data were checked for node points, member designations, member properties, loads and load combinations, seismic factors, member releases, joint releases, joint connections, etc. The output data were also reviewed for reliability of result A member and stiffener plate were added for reinforcement of this ancho All support calculations and justifications were acceptable except for problems affecting Anchor 1E51-41A53 as stated below:
During the support walkdown reinspection (see next paragraph), the inspectors noted that at least three other pipe supports, one for a 6" diameter pipe and two for 4" diameter pipe, were attached to Civil Structural Beams which also supported Anchor 1E51-41A53, and were used to qualify the anchor and its deflection requirement. The loads from the additional attachments could affect the analysis and the deflections of the civil structural beams and anchor 1E51-41A53, but were not considered in the STRUDL analysis. This problem was discussed with the licensee engineers and management during the exit meeting. CP&L management, at the exit meeting, agreed to perform reanalysis to add the loads from the other attachment Pending the licensee resolution of this support calculation, this item is identified as Inspector Follow-up Item (IFI) 50-325/89-15-01, Pipe Support Calculation Problem for Mark No. 1E51-41A5 _ - _ - _ _ _ - _ _ _ ___ _
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l gg 8 The 12 support calculations reviewed are listed below:
Mark N Node No.-
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1E51-3PG49 395 1G31-50SS9 550 1G31-50SS8 543 1E51-3VH48 374 1G31-14VH7 524 .
1G31-50VH10 554 4
- 1E51-41FH51 425 1E51-3SS46 370 E51-3SS47 372 4 1E51-41FH50 415 1E51-41PG52 430 1E51-41A53 470 Pipe Support Walkdown Reinspection
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In order to assure the adequacy of the modification on this system, the inspectors walked down the system, with assistance from the licensee engineer and QC inspectors, to check the moditied piping, new and existing valves, pipe supports near the modified piping, and modified and. relocated support Fabrication Isometrics SK-M-89-053-010, Rev. A, and 011, Rev. A of. Dwg. No. FSP-25098 and Analysis Isometric Sheet No. 563, Dw No. F-28046, Rev. 3, were used to check pipe diameter, dimensions, new valve location, valve identification, drain-test branch locations, and weld qualitie Eight pipe supports were reinspected. Support No s . 1E51-41SS 51, 1E51-41S S52, 1G31-14VH7, 1G31-50SS8, 1G31-50SS9, and IG31-50VH10 were checked against the as-built drawings for general configuration, spring settings. snubber liquid levels and damage, deformation, loose or missing parts, etc. Support No.1E51-41FH50, Rev. A., was checked for the relocated dimension, weld sizes, rod sizes, i and member size Support No. 1E51-41A53, Rev. A (Anchor) wa.< checked for l the site reinforcing member and stiffener plate, wald r.izes, weld configuration, and sizes of civil structural beams. The civil structural beams were used in the analysis for anchor qualification. The inspectors found two discrepancies during the anchor reinspection: First, Support No. PS-3896 in the vicinity was found to have a loose rod on one side and
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a bent rod on the other sid The licensee engineer stated that this support is a nonsafety related support. But the licensee engineer agreed to review the impact of its failure on the safety-related system .
Second, at least three other pipe supports, one for 6" diameter pipe and l two for 4" diameter pipe, were found to be attached to the civil structure i beams which were used to qualify Anchor No 1E51-41A5 The loads from those additional supports were not considered in the computer analysi ;
All other reinspected supports were accepte '
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5. Exit Interview The inspection scope and results were summarized on June 23, 1989, with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection results listed belo Proprietary information is not contained in this repor Dissenting comments were not received from the license /89-15-01 IFI, Pipe Support Calculation Problem for Mark No. 1E51-41A53 325/89-15-02 IFI, Radiographic Indication in Weld 251 Station (45).
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