IR 05000259/1993021
| ML18036B319 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/27/1993 |
| From: | Blake J, Kleinsorge W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18036B318 | List: |
| References | |
| 50-259-93-21, 50-260-93-21, 50-296-93-21, NUDOCS 9306160113 | |
| Download: ML18036B319 (12) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/93-21, 50-260/93-21, and 50-296/93-21 Licensee:
Tennessee Valley Authority 3B Lookout Place 1101 Harket Street Chattanooga, TN 37402-2801 Docket Nos.:
50-259, 50-260 and 50-296 License Nos.:
DPR-'33, DPR-52, and DPR-68 Facility Name:
Browns Ferry 1, 2, and
Inspection Conduct
17-21, 1993 Inspector:
W.
K einsorge P.E.
Reactor Inspector D te igned
Approved byJ. J.
Bla e, Chief Hat ial and Process Section En ineering Branch Division, of Reactor Safety
+rr7 Date Signed SUMHARY Scope:
This routine, announced inspection was conducted in the areas of Flow Accelerated Corrosion (FAC), Hicrobiologically Induced Corrosion (HIC), safety relief valve testing, weld material control and followup on the licensees actions on previously identified enforcement items.
Results:
It appears that the licensee's FAC program will be an effective tool in maintaining high energy carbon steel piping systems within acceptable wall thickness limits.
A weakness was identified related to the licensee's lack of:
a formal documented radiographic HIC indication acceptance criteria; calculations providing the basis for that acceptance criteria; and a formal transmittal of that HIC radiographic acceptance criteria.
(Paragraph 3)
The licensee's control of welding filler material is consistent with procedural and regulatory requirements.
In the areas inspected, no violations or deviations were identified.
930hib01$ 3 '730601 PDR ADOCK 05000259
REPORT DETAILS Persons Contacted Licensee Employees:
- H. Bajestani, Techn'ical Support Manager
- J. Corey, Site Radiological Control Manager
- F. Fiosiello, ISI/gA
- K. Grooms, Welding Supervisor Hods
- H. Herrell, Operations Manager
- J. Sabades, Superintendent, Chemistry
- P. 'Salas, Licensing Manager Other licensee employees or contractors contacted. included craftsmen, technicians, engineers, and office personnel.
NRC Personnel:
- C. Patterson, Senior Resident Inspector
- J. Hunday, Resident Inspector
. *G. Schnebli, Resident Inspector
- T. Liu, Intern
- Attended exit interview 2.
Acronyms and initialisms used throughout this report are listed in the last paragraph.
Flow Accelerated Corrosion (49001)
The licensee has established a Flow Accelerated Corrosion (FAC)
inspection program, for Unit 2 only, which uses the CHECHATE EPRI (Electric Power Research Institute) computer code, industry experience, and previous inspection data as predictive tools for determining and prioritizing inspection locations.
The licensee's Unit 2 CHECHATE electronic model includes the systems that have been predicted to be susceptible to FAC, which ranks the systems components according to their FAC susceptibility.
The licensee's primary basis for the selection of components for ultrasonic (UT)
examination during the next refueling outage is engineering judgement including plant specific design characteristics.
In addition to engineering judgement, the licensee uses:
the results of previous ultrasonic examinations; predictive analyses; and industry and operating experience.
The licensee program has criteria for excluding a system from the FAC program.
The licensee's FAC program includes the following systems:
Feedwater; Condensate; Feedwater Heater Drains; Moisture Separator. Drains; Extraction Steam; Hain Steam Drains; Auxiliary Steam; and Crossover and Crossunder piping.
Components in the licensee's program include: control/check valves; tees and branches; expanders and reducers; flow nozzles and orifices; exit nozzles; straight pipe sections; crosses; and elbows and reducing elbow The licensee's procedure for performing UT thickness examinations of carbon steel piping is approved by the management review committee defined in Section 6.0 of the licensee's Technical Specifications (Plant Operations Review Committee).
The licensee's procedures define the method for the following:
performing UT measurements; grid marking the inspection zone; ensuring location repeatability on followup UT examinations; ensuring certified UT examiners perform the examinations; and establishing acceptance criteria for evaluating UT data.
The licensee program provides guidance only for analyzing results of UT measurements (there are no firm requirements in this area).
The criteria for establishing minimum wall thickness (t.) requirements for safety-related (SR) piping is to set t. to code minimum allowable thickness.
The criteria for establishing t. requirements for Balance of Plant (BOP) piping is to set t. to code minimum allowable thickness.
The wear rate (wr) calculation method is determined by the point to point method.
The predicted thickness (t~)-[t,,~-(wr x current operating cycle time)] for previously measured components and t~=[t ~-(wr x total operating time to date)] for first time examined components.
Component acceptability at the next outage is determined by the following method:
t,,~ > t and t,~
> t.
The licensee's method for calculating remaining life rl) = (t,,~-t )/wr.
To date no repairs or replacements have been made to ASHE Code class 1,
2, or 3 components identified by the FAC program.
The licensee has a
repair and replacement program for BOP components which fail to meet the licensee's FAC acceptance criteria.
It should be noted that the program is not contained in a single document.
Welding and nondestructive examination (NDE) of SR and BOP piping replacements are being accomplished by qualified personnel and suitable procedures.
The Corporate Nuclear Engineering Department is responsible for the implementation of the FAC program.
The licensee does not have formal programmatic requirements assigning the responsibility for oversight of the FAC program.
The licensee's FAC program for ASHE Class 1, 2, and
components is covered by their guality Assurance (gA) program.
To evaluate the licensee's performance in this area the inspector performed a walkdown inspection of pipe grid locations, observed overlay
'eld repair to two tee connections, conducted interviews with licensee and contractor personnel, and reviewed Work Order (WO) package 93-06174-00.
The inspector reviewed welding material. certification documentation, and welder qualification, and qualification maintenance records associated with the above WO.
In addition the inspector reviewed the below listed procedure Identification Revision G-97B 5/5/91 NIH-004 4/30/92 2-TI-140 3/2/93 2-TI-148 3/2/93 Title General Engineering Specification, Corrosion Control; Part B, Erosion-Corrosion Project Engineering Specification, Erosion/Corrosion Technical Instruction, Pipe Wall Degradation Honitoring Program for Dual Phase Fluid Systems Technical Instruction, Pipe wall Degradation Honitoring Program for Single Phase Fluid Systems The grid spacing criteria specified in G-97B is less restrictive than that recommended by EPRI.
The grid spacing in G-97B is based
'on the recommendations promulgated in ASHE BEPV Code Case N-480, which predates the current EPRI recommendations.
It should be noted that the implementing procedures, 2-TI-140 and 2-TI-148, are consistent with the current EPRI recommendations.
The licensee indicated that there are two locations where liquid sealant had been used to temporarily correct leaks caused by FAC.
Both have been corrected by pipe replacement with only post maintenance testing remaining to close out the work orders.
The maintenance planners indicated that the permanent repair requirement prior to close out is placed in all liquid sealant repair Work Orders resulting from FAC.
G-97B, paragraph 4.8. 1.3, erroneously requires ultrasonic thickness readings to be taken at the center of grid squares instead of at the grid intersections.
The licensee indicated that they were aware of this error and would take steps to effect its correction.
The UT data collected during the current outage (Refueling Outage 6) is the first data collected for CHECHATE~.
During this outage the licensee planned to examine 355 components which expanded to 358 components.
Five components were replaced.
The expansion was appropriate.
With this data the l.icensee intends to implement the T-Dat portion of the CHECHATE program.
It appears that the licensee's FAC program will be an effective tool in maintaining high energy carbon steel piping systems within acceptable wall thickness limits.
Within the areas examined no violations or deviations were identifie ~
s.
Examination of EECW Welds for HIC (62700)
By letter dated September 29, 1988, TVA committed to the NRC that a
population of the Unit 2 Emergency Equipment Cooling Water (EECW) butt welds which were previously inspected by radiography (RT) will be re-radiographed before Unit 2 restart and at each subsequent Unit 2 outage, to ensure structural integrity of the system.
Any increase in indications will be re-analyzed to determine any affect on structural integrity, of the system.
The licensee RT examined 37 welds of which nine had previously been RT examined.
Of the nine, four had previously (1987/8) exhibited,HIC indications in or adjacent to the welds, the remaining five exhibited no HIC indications when examined in 1987/8.
Two of the nine previously RT examined welds exhibited HIC when examined in 1987/8, but exhibited no.
growth of the HIC indications lengths, when reexamined in 1992.
In addition to the nine previously examined welds, discussed above, the licensee selected 29 additional welds to RT examine.
Of the 29, six exhibited HIC indications.
Of the total population of 37 welds
exhibited HIC indications.
The inspector conducted interviews with licensee personnel, and examined the radiographs for eight welds listed below.
Based on the eight film packages examined and,the interviews the inspector considers the sample
'selection, radiographic film quality, and interpretation acceptable to determine the extent of HIC growth.
Weld Radiographs Examined T-EECW-2-CAC-07 T-EECW-2-AC-23 T-EECW-2-BD-25 T-EECW-2-CAC-25 T-EECW-2-BD-13 T-EECW-2-BD-31 T-EECW-2-AC-19 T-EECW-2-BD-14 The inspector requested the HIC indications RT acceptance criteria, and the calculations providing the basis for that acceptance criteria, on
~ Honday, Hay 17, 1993, the first day of the inspection.
On Friday, Hay 21, 1993, 15 minutes prior to the exit interview, the licensee provided the inspector with Civil Design Standard DS-C1.2.8,
"Special Requirements Structural Evaluation of Hicrobiological Influenced Corrosion Degradation in Piping" Revision 1.
One hour and thirty minutes after the exit interview the inspector was contacted in a conference call by the licensee's HIC expert who identified which approach, of the many
'contained in the DS-C1.2.8, that was used to determine the bounding flaw-size (the HIC RT acceptance criteria).
The licensee admitted that the calculation had been informal and undocumented; the bounding flaw size was verbally transmitted to the site as a percentage of the circumference of the pipe.
After review of DS-C1.2.8 and the aggregate length of the HIC indications in each weld, the inspector, concurs with the licensee, that the EECW
system in Unit 2 is acceptable for continued operations for fuel cycle seven.
The inspector considers the licensee's responsiveness in this issue poor.
The lack of 1)
a formal documented HIC indication RT acceptance criteria; 2) calculations providing the basis for that RT acceptance criteria; and 3)
a formal transmittal of that HIC indication RT acceptance criteria is considered a weakness.
This issue is further discussed in NRC Report 50-259,260,296/93-18.
Within the areas examined no violations or deviations were identified.
Safety Relief Valve (SRV) Testing (73756)
During performance of the SI to test the instrument line excess flow check valve operability on Hay ll, 1993, a high reactor pressure condition resulted in an ATWS/ARI/RPT trip of both recirculation pumps, a
depressurization of the scram pilot air header and a subsequent scram condition.
This SI was being performed in conjunction with the reactor pressure vessel system leak test, 2-SI-3.3.1.A, ASME BKPV Code Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping (ASHE Section III, Class 1).
The primary reactor pressure instrument being used to control reactor pressure for the leak test was unknowingly valved out of service as required by the Harotta valve testing.
The ATWAS/ATI/RPT trip occurred at 1118 PSIG, SRV 1-34 set point is 1105 PSIG.
To determine why the SRV did not release at its setpoint of 1105 PSIG the licensee had General Electric Company (GE) perform an evaluation of the incident and conducted a test of the SRV pilot.
The inspector examined the engineering evaluation and witnessed the SRV pilot testing.
Because the media for the pressure test was 199 'F water, and not 550 'F steam, the valve would be expected to perform differently.
The spring in the SRV pilot will be stiffer at 199 'F than at 550 'F, causing the effective set point to rise.
The expected pressure area on the valve disk in the pilot would be smaller for a liquid than for a gas because the gas will flow further into the seating zone.
Thus more pressure would be required on a smaller area with the liquid media water, which would cause the effective set point to rise.
The evaluation indicated that the expected set point with a 199 'F water media should be 3.6% higher (at or above 1120 PSIG).
The pilot was tested at the Wile Laboratories, Huntsville AL facility.
The pilot was tested five times with the following results:
1100, 1103, 1107, 1103, and 1105 PSIG release points.
The acceptance criteria is 1105 i 11 PSIG.
Clearly the valve
"
will operate as required in its intended environment.
This issue is further discussed in NRC Report 50-259,260,296/93-18.
Within the area examined no violations or deviations were identifie The inspector reviewed procedures, conducted interviews with licensee and contractor personnel and inspected both of the site weld material issue stations to determine whether the control of weld filler material was in accordance with procedural and regulatory requirements.
The licensee has a dual system for the control of weld filler materials.
Meld filler material used by GE is controlled by the GE gA program, -implemented by TVA approved, GE procedure GE-TVA-21.0, "Weld Haterial Issue, Storage and Control", Revision 0.
All other weld filler material is controlled by the TVA gA program implemented by TVA procedure SSP-7.51,
"Controlling Welding Brazing, and Soldering (WBS) Haterials",
Revision 2.
The TVA and GE weld filler material control programs are very similar, in that both require complete accountability of welding filler material while the material is in the possession of the welder (from the point of issue to the point of return).
Each piece of weld filler material (electrode or rod) issued to the welder must be returned and one-for-one accounted for as an unused piece, damaged piece, or a stub.
The programs differ in the treatment of the returned material.
The TVA program assumes that usable returned material will be reused, and therefore requires that the unused pieces, damaged pieces, and the stubs be returned to the welding material issue stations for an accounting.
(The only exception to this requirement is made for weld material that has become radiologically contaminated a'nd must be disposed of as radioactive waste.
In this case the welder must get a receipt from Radiological Control which accounts for each piece so disposed.)
The GE program assumes that all weld filler material issued to welders has been radiologically contaminated and is disposed of as radioactive waste after a one-for-one accounting of each returned unused piece, damaged piece or the stub of weld filler material.
This accounting is done inside the contamination zone whereupon the unused pieces, damaged pieces and the stubs are put through a narrow slot in the top of a locked 55 gallon drum.
When the drum becomes full or GE deactivates the site, the drum, with lock intact, is turned over to Radiological Control for disposal as contaminated waste.
Both programs provide a reasonable level of assurance that returned weld filler material will not be used in unauthorized applications.
During the inspection of the weld material issue stations, the inspector noted a
memo in the drawer containing uncoated welding rods, placing a
hold on the issue of all 1/8-inch diameter, Type 70S-3 material.
The licensee representative informed the inspector that the hold status was an administrative control to prevent a recurrence of the circumstances that resulted in NRC Violation 50-260/93-05-01, issued for failure of a welder to follow welding procedure specification requirements.
The licensee further stated that this administrative control was not included as part of the corrective actions committed to by the licensee,.in their letter of April 21, 1993, in response to the violation.
Therefore, the hold status could be lifted by the welding engineer who imposed it without prior notification to the NRC.
The inspector reviewed NRC inspection report 50-259,260,296/93-06 dated Harch 23, 1993 and TVA
Letter dated April 21, 1993 and finds the licensee actions consistent with regulatory requirements.
The licensee's control of welding filler material is consistent with procedural and regulatory requirements.
Within the areas examined, no violations or deviations were identified.
Followup on licensee's Action on'Previously Identified Enforcement Hatters (92702)
(Closed) Violation 50-259,260,296/93-22-01:
"Failure to Implement Changes to a Controlled Document" This violation contains multiple examples of failure to implement changes to the General Construction Specification G-29.
The NRC has reviewed the licensee's letter of response dated August 5, 1992, and found it acceptable.
In their letter, the licensee admitted to the violation and attributed the cause of this violation to -inadequate filing instructions.
The licensee updated the documents not previously updated, assigned an individual to perform receipt verification of corporate documents; examined 106 work plans for the effects of a lack of updating G-29, performed an assessment G-29 documents at two other TVA facilities, revised corporate procedure NEP-5.5, and reviewed and corrected the table of contents to G-29.
The inspector examined the licensee's corrective actions as described in their letter of response, and determined that the licensee has taken appropriate actions to prevent recurrence.
Th'is matter will be closed.
With in the areas examined, no violations or deviations-were identified.
Exit Interview The inspection scope and results were summarized on Hay 21, 1993, with those persons indicated in paragraph 1.
The inspector described the areas inspected.
Although reviewed during this inspection, proprietary information is not contained in this report.
No dissenting comments were received from the licensee.
Acronyms and Initialisms ASHE BOP BKPV DPR EECW EPRI FAC GE HIC NDE No.
American Society of Hechanical Engineers Balance of Plant Boiler and Pressure Vessel
,Demonstration Power Reactor Emergency Equipment Cooling Water Electric Power Research Institute Flow Accelerated Corrosion General Electric Hicrobiologically Induced Corrosion Nondestructive Examination Number
~
'
Acronyms and Initialisms Cont'd NRC P.E.
PSIG gArl RT SR SRV tmeeeured t
.n tnominel tprea TVA UT WBS WO wr Nuclear Regulatory Commission Professional Engineer Pounds per Square Inch Gauge equality Assurance Remaining Life Radiography Safety Related Safety relief Valve Heasured Thickness Hinimum Wall Thickness Nominal Thickness Predicted Thickness Tennessee Valley Authority Ultrasonic Welding Brazing and Soldering Work Order Wear Rate