IR 05000324/1990009
| ML20033G816 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/28/1990 |
| From: | Belisle G, Whitener H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20033G814 | List: |
| References | |
| 50-324-90-09, 50-324-90-9, 50-325-90-09, 50-325-90-9, NUDOCS 9004120167 | |
| Download: ML20033G816 (15) | |
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p* #80 UNIT ED ST ATES .8 0g#o - NUCLEAR C.ECULATlRY COMMIT $10N , , [" C EoloN li ' o g 1o1 MARIETTA ST RE ET, N.W.
- ATLANT A, GEoRot A 30323
%,*e,,,s Report Nos.: 50-325/90-09 and 50-324/90-09 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 ' Docket Nos.: 50-325/9D-09 and 50-324/90-09 License Nos.: DPR-71 and DPR-62 . Facility Name: Brunswick 1 and 2 laspection Conducted: February 13-20, 1990 Inspectors: N.~i,//kr_Ezv 8-29-76
~ < H. L. Whi tene'r Date Signed Accompanying Inspectors: P. A, Taylor J. J. Lenahan Approved by: ESC C8 2 21
- /O G. A. Belisle, Chief Da'te Signed
' Test Programs Section Engineering Branch Division of Reactor Safety SUMMARY Scope: This routine, announced inspection was conducted in the areas of witnessing and evaluating the containment intergrated leak rate -test (CILRT) on Unit 2, reviewing improvements to the local leak rate test program, witnessing and eeluating the hydrostatic test of replacement piping for the reactor vessel recirculation loops, and followup of previous inspection findings and licensee oyent reports.
Results: In the areas inspected, violations or deviations were not identified.
Within the areas of hydrostatic testing and integrated leak rate surveillance testing, the inspectors found that test procedures were acceptable and that personnel were knowledgeable in their areas of performance.
In the area of local leak rate testing, the inspectors observed that the licensee had taken positive action to improve the quality, of the test program including: plant modifications, more attention on root cause analysis, and computerizing valve performance history. Management involvement is evident in these steps, especially in regard to plant modifications.
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Persons Cor', acted Licensee Employees .
- C.
Blackmon, Manager, Operations
- M.
Blinson, Engineer, Inservice Inspection H.
Bordeaux, Supervisor, Quality Control J.
Crider, Engineer,-Inservice Inspection-
- W.
Norman, Manager, Quality Assurance / Quality Control P.
Godsey, Inservice Inspection Specialist
- J.
Harness, Plant General Manager (PGM)
- A.
Hegler,.Radwaste/ Fire Protection Supervisor
- R.
Helme, Manager, Technical Support
- J.
Moyer, Technical Assistance to PGM
- R.
Poulk, Supervisor, Regulatory Compliance D.
Savage, Outage Ccordinator L.
Wheatly, Supervisor, Inservice Inspection Other 11 cense'e employees contacted during this inspection included ' engineers, operators, technicians, and administrative personnel.
Other Organizations J. Blessing, United Energy Services R. Shrik, United Energy Services (NRC) Resident Inspectors
- W.
Ruland, Senior Resident Inspector W.
Levis, Resident Inspector D.
Nelson, Resident Inspector
- Attended exit interview 2.
Observation of Hydrostatic Testing for Unit 2 Recirculation. Piping Replacement / Modification (72701) The inspector examined procedures, work activities, and reviewed. test data.
relating to the Unit 2 reactor vessel hydrostatic test. This test was performed to pressure test the recirculation piping which had been replaced during the current refueling outage.
Acceptance criteria utilized by the inspector appear in Technical Specification 4.0.5 and the American Society of Mechanical Engineers (ASME) Code, Section XI,1980 edition through the Winter of 1981 Addenda.
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Review of Hydrostatic Test Procedures
The inspector examined Periodic Test Procedure PT-80.1, Reactor Pressure Vessel Hydrostatic Test.
This procedure specifies the
requirement for hydrostatic testing the reactor vessel pressure ' boundary once each 10 year inspection interval, or whenever repair, replacement, or a system alteration is performed.
The inspector verified that test initial conditions and prerequisites were , specified, test intructions and objectives were clearly stated, and ' test acceptance criteria were specified.
The hydrostatic test pressure specified was in accordance with ASME Section XI, Article l IWB 5000.
The test condition holding time in the test procedure complied with ASME Section XI, Article IWA-5213, a four hour holding time af ter attaining test pressure for insulated systems, and ten minutes for noninsulated systems or components.
The procedure specified the hydrostatic test boundary, and contained precautions to assure that isolation protection'was maintained between high and. low pressure systems to avoid overpressuring the low pressure systems ' (residual heat removal and core spray). The inspector also examined the following procedures referenced by PT 80.1 which contained additional hydrostatic test requirements: ENP-16, Procedure for Administrative Control of Inservice Inspection Activities OCM-SUP 501, Pinning Support Hangers for MS, PSN, HPCI, RCIC and ' RHR Systems PT 80.2, Clast 1 Conditional System Leakage Test I b.
Observation of Hydrostatic Test The inspector witnessed the start of the reactor vessel cnd system pressurization from the control room.
Pressure increases were less than 50 psi per minute in accordance with procedure PT 80.1 precauticns. The inspector verified that the residual heat removal (RHR) and core spray (CS) systems were not being pressurized.
Licensee operators monitored pressure instrument numbers E11-PIC-606A, E11-PIC-606B, E21-R-600A and E21-R-600B to verify that the RHR and CS systems were not being pressurized. The inspector was present in the control room when system test pressure was obtained.
PT 80.1 specifies that the test pressure be greater than 1088 psig but less than 1103 psig.
After some period of minor pressure - fluctation, the system pressure stabilized at approximately 1097 psig ! and the hold periods (a minimum of four hours for insulated lines and ten minutes for noninsulated lines) were maintained prior to commencing visual examinations of the lines.
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. The inspector accompanied licensee QC inspection-personnel during walk down inspections conducted four hours or more after pressure was maintained on the reactor coolant pressure boundary piping. The inspector examined various piping systems within the hydrostatic test boundary, including the new recirculation piping system piping. No through wall piping leaks were observed.
Some leakage from mechanical joints, gaskets, and valve packing was observed.
These leaks were documented on the Leak Identification Data Sheets - (Attachment 1 to PT 80.1).
c.
Review of Test Data The inspector reviewed portions of procedures PT 80.1 and OCM-SUP.501 which had been completed prior to the start of pressurization (test prerequisites and initial conditions) and during the inspection The s inspector verified that test data had been accurately recorded and that deviations from test procedures-were documented for review and evaluation.
The inspector reviewed work requests that had been written to document and repair flange and stem packing leaks which occurred daring the hydrostatic test.
After repairs are completed, the repaired areas will be reinspected during startup when the reactor-coolant ' system is at pressure in accordance with PT 80.2.
The inspector also examined a liquid penetrant non destructive examination (NDE) report which ' documents the NDE examination performed on line number 2-B2-708, a lh inch pipe which is one of.the reactor level recorder / instrumentation lines. This examination was performed at the request of the Resident Inspector to determine if the line had been damaged during the outage since it appeared that this pipe may have been used to support a wire rope sling used in a lifting operation. The resident inspector noted some moisture on this line approximately 24 hours af ter the hydrostatic test had bcen
started. The pressure in the system at this time was approximately 950 psig.
The results of the liquid penetrant examination disclosed no indications.
The licensee stated that they would examine this line again when the system was at pressure after startup.
Within the are.as inspected, no violations or deviations were identified.
I 3.
Containment Integrated Leak Rate Test - (70307)(70313) (Unit 2) The inspector reviewed and witnessed test activities to determine that the I primary containment integrated leak rate test was performed in accordance with the requirements of Appendix J to 10 CFR 50, Primary Reactor _ Containment Leakage Testing for WLter-Cooled Power Reactors; ANS1 N45.4-1972, Leakage-Rate Testing of Containment Structures for Nuclear ' Reactors: BN-TOP-1, Revision 1-1972, Testing Criteria For Integrated i Leakage Rate Testing of Primary Containment Structures For Nuclear Power Plants; and, test procedure PT-20.52 Integrated Primary Containment Leak Rate lest.
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. Selected sampling of the licensee's activities which were inspected included: (1) reviewing the test procedure to verify th at it was properly, approved and conformed with regulatory requirements; (2) observing the test performance to determine that test prerequis',tes were completed, special ' equipment was installed, instrumentation was calibrated, and appropriate data were recorded; and (3) evaluating the preliminary leak rate data to verify that leak rate limits were met. Pertinent aspects are discussed in the following. paragraphs: a.
General Observations The inspector witnessed and reviewed portions of the test preparation, containment pressurization, ter.perature stabilization, and data processing during the period from February 13-20, 1990, and concluded the following: (1) The test was conducted in accordance with an approved procedure and procedure changes and test discrepancies were documented.
(2) Selected test prerequisites were found to be. completed.
(3) Plant. systems required to maintain test control were found to be operational.
(4) Special test instrumentation was found to be _ installed and calibrated.
(5) Data required for performing the l containment leak rate calculations were recorded at 15-minute intervals.
(6) Problems encountered during the test were described in the test event log.
(7) Pressurized gas sources were properly isolated and vented to preclude in-leakage or interference of out-leakage through containment isolation valves.
(8) Selected procedure valve alignments were reviewed against system drawings to verify correct boundary alignment, and venting and draining of specific systems.
(9) Temperature, pressure, dew point, and flow data were recorded at i 15-minute intervals. Data were assembled and retained for final ! evaluation and analysis by the licensee.
A final' Integrated i Leak Rate Test (ILRT) report will be submitted to the Office of i Nuclear Reactor Regulation, i ! I.
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Procedure Review - Unit I and Unit 2 (70307) , , Portions of PT-20.5, Revision 20, were reviewed to verify that adequate test controls, instructions, and acceptance criteria 'were specified.
Additionally, limited valve alignments and venting and- , draining instructions were reviewed.- Additional test procedures and' associated documentation reviewed totally or, in part, during this inspection included: " (1) PT-20.3, Revision 29 Local Leak Rate Testing The licensee is in the process of revising this procedure to incorporate specific draining instructions.
(2) ENP-16.4, Revision 8, Use_of Leak Rate Test Equipment This procedure provides generic instructions for the various Type B and Type C test methods.
(3) ENP-16.8,_ Revision 1 Containment Leakage Tracking -- This procedure provides instructions for. maintaining a. local leat rate test log which specifies the running total leakage for Type B and Type C leak rate tests.
(4) PT-20.5.1, dated 2/17/88, Primary Containment Inspection-This procedure provides specific instructions and requirements for inspecting the primary containment for deteriorations or structural damage.
(5) Containment Running Total Log for Type B and Type C leakage during the 1989-90 refueling outage.
The inspector determined that the log was current and total leakage was less than 0.6 La.
(6) Unit 2 local leak rate status f'or 1989-90 refueling outage This data included as found and e,s lef t -leak rates, plant - modifications, and work orders.
I (7) Valve Performance History This is a computerized history of valve leakage.
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Test Performance - Unit 2 (70313) , (1) Test Method
The licensee utilized the integrated leak rate analysis program i of United Energy Services which had data analysis capability for j the total time analysis in accordance with requirements of BN-TOP-1 for a short duration test and mass point-linear { regression analysis in accordance with the recommendations of
ANSI /ANS-N56.8-1981, Containment System Leakage. Testing Requirments, for a 24 hour test.
For the Unit 2 test, the data ! met the BN-TOP-1 criteria. Consequently, a short duration test ' of eight hours and a verification test of four hours were . i performed using the total time analysis method stated in BN-TOP-1
- i (2) Test Description l Pressurization of the containment was initiated,at 11:35 a.m. on February 16, 1990,- and was terminated at approximately 8:04 p.m.
The four hour stabilization period was initiated at this time
! and the Type A test initial data point was taken at 1L34 a.m.
.; ! on February 17. The initial test was terminated at 7:00 a.m. on ! i February 17 with a measured leak rate of 0.55 wt. percent per day l which exceeded the allowable leak rate limit of 0.375 wt.
percent per day (0.75 La).
A search for the leakage' path was , initiated.
In addition to several minor leaks, a major leak was identified in the nitrogen inerting header.
Leakage through the inerting t header was isolated by using a local leak rate test rig to supply regulated air at a pr6ssure less than containment pressure to back pressurize this leakage path. Pressure in the header was monitored throughout the test' to ensure pressure remained less than containment pressure.
Isolation of the nitrogen inerting header reduced the leak rate from 0.55.to 0.4 wt percent per day, which is'still in excess of 0.75 La.
, The search for containment leakage was continued and a second large leak path was identified through thF reactor building to . torus vacuum breaker valves, CAC-V16 and CAC-X20A. Leakage past the inboard butterfly valve, CAC-V16, was escaping past the " outboard check valve CAC-X20A, without a pressure buildup in the volume between the valves.
It was concluded that the check , valve was not properly seated. To resolve this leakage path the inboard valve CAC-V16 was opened to allow ' full containment pressure against the check valve CAC-X20A.
This apparently seated the check valve and reduced the leak rate from 0.4 to 0.27 wt. percent per day.
The Type A test was re-started at . 4:19 p.m.
on February 18 and concluded at 12:19 a.m.
February 19.
Following a one hour period of stabilization, a four hour verification test was successfully completed at , , , % _ ~ . .. .
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. 5:34 a.m. on February 19, 1990. The following table summarizes identified leakage and action taken to reduce the-leak rate.
Leak Path Leakage Components Action 'E11-F021A unknown packing adjusted E41-F078 unknown draincap' none - TIP System unknown pipe plug none 1 scfh fitting none 2 sefh tubing none Gas monitor-unknown fittings none Inerting Header 80 sefh CAC-V4, V5 pressurized V6, V15 ~ to isolate Torus to SBGT 10 scfh CAC-V7, V8: pressurized, - to isolate Drywell to SBGT 4 scfh CAC-V9, V10 pressurized.
to isolate Unktmwn 1 inch Reactor Vessel Refill Vessel Water /hr and pump the , sumps prior to starting the test.
Reactor 70 sefh CAC-V16,X20A Opened inside Building To Torus valve to Vacuum Breakers seat outside check valve' 3.
Test Results - Unit 2 (a) Type A Test Technical Specification allowable leakage (La) for. Brunswick-Unit 2 is 0.5 wt. percent per day. Therefore, the integrated leak rate test leakage limit of 0.75 La as required by Appendix J.is 0.375 wt.
percent per day.
In the period following containment stabilization.
I the containment leakage was approximately 0.55 wt. percent per day which exceeded the test acceptance limit.
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j . ' . After isolating leakage through the nitorgen inerting header, purge to the $BGT, and reactor building to torus vacuum breakers, the leak rate was within the allowable leakage rate.
The following are the results of the leak rate measurement and the leak rate compensated for error (95 percent Upper Confidence Limit) for both the Mass Point (MP) analysis according to ANSI /ANS 56.8 and total time (TT); analysis according to BN-TOP-1: Measured Leak Rate 95 % Upper ' Confidence Limit MP 0.278 wt. percent per day 0.281 wt. percent per day TT 0.270 wt. percent per day 0.308 wt. percent per day The inspectors calculated weighted averages for containment temperature, pressure, and vapor pressure using the weighting factors and individual sensor data for a sample of data sets to verify-agreement with the weighted averages generated by the licensee's computer program.
Subsequently, the weighted averages generated by the licensee's-program were used by the inspector.to calculate mass, leak rate, and the 95 percent upper confidence leak rate.
The inspector's calculations agreed with the licensee's calculations.
(b) Supplemental Test - Unit 2 Appendix J requires that a supplemental test be performed to verify the accuracy of the Type A test and the ability of the containment ILRT instrumentation to measure a change'in. leak rate. The following is an acceptable supplemental test method as described in Appendix C of ANSI N45.4 - 1972.
A know leak rate (Lo) is imposed on the containment and the measured composite leak rate -(Lc) must equal, within 10.25 La, the sum of the measured leak rate (Lam) plus the known leak rate (Lo).
The acceptance criteria is expressed as: Lo + Lam - 0.25 La < Lc < Lo + Lam + 0.25 La A 4 hour verification test was performed.
The following measured values were obtained (Units are in wt. percent per day): TT MP LAM 0.270 0.278 . ' i <
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' TT MP Lo 0.505 0.505 Le 0.753 0.752 .25 La 0.125 0.125 .- Substitution of these values into the acceptance criteria shows that the inequality equation was satisfied as follows: TT 0.65 < 0.753 < 0.90 MP 0.658 < 0.752 < 0.908 The inspector concluded that the verification test confirmed the instrument system capability to measure the containment leak rate, d.
Type A Test Status-Unit 2 The initial containment leak rate was - greater than the test acceptance limit defined as 0.75 La in 10 CFR 50, Appendix J.
Paragraph III.A.5(b).
In accordance with Appendix J, Paragraphs III. A.1(a) and III. A.6(b), the test was identified as a failed test.
Paragraph III. A.o(b) requires that if two consecutive periodic Type A tests fail to meet the applicable acceptance criteria in III.A.5(b), notwithstanding the periodic retest schedule of III.D a Type A test shall be performed at each plant shutdown for refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the acceptance criteria in III. A.5(b), af ter which time the retest schedule specified in III.D may be resumed.
Prior to this test Unit 2 was on the accelerated integrated leak rate test schedule of Paragraph III.A.6(b) as a result of consecutive Type A test failures. With the failure of this Type A test the licensee has not yet performed two consecutive successful Type A tests on Unit 2; therefore, Unit 2 remains on the accelerated Type A test schedule.
4.
Additional Inspection Related to Leak Rate Testing (Unit 2) 70313 a.
As Found Leak Rate q In a preliminary evaluation of leakage corrected in the local leak , rate test program, the licensee estimated by minimum path leak rate calculation that containment leakage was reduced by about 100 scfh prior to the Type A test.
For a volume of 294,981 cubic feet and accident pressure (Pa) of 49 psig, 100 scfh equates'to 0.19 wt. percent per day. This value when added to the final measured leak rate of '~ 0.27 wt.
percent per day, yields an "as found" leak rate of 0.46 wt.
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The value 0.46 wt. percent per day exceeds the acceptance limit of 0.375 wt. percent per day and indicates a failed.
"as found" leak rate test.. However,'these calculations are based on preliminary data.
A final : evaluation will be included in the leak J rate test report.
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Local Leak Rate Program (Unit 2) 61720 To gain a better understanding of the -status of the local leak rate - [ test program the inspectors discussed action,taken or ' planned to
upgrade the' program with licensee personnel.
Areas ' discussed-j included the following: ,
(1) Plant Modifications (PMs) Five PMs were completed in 'the 1989/90 refueling outage.
The
- general effect of the PMs include:
i i Permit local testing in the accident direction,
! Permit local testing of individual valves, and
! Upgrade valve design.
! Plant modification packages were not reviewed during this [ inspection; however, retest after modification was indicated in , the local leak rate test status.
(2) Root Cause Analysis
The licensee has established a Component Engineering Group which l has responsibility for performing root cause analysis for valve j and other component failures.
The inspector reviewed-preliminary root cause analysis of ' the drywell head gasket' failure and concluded that the. analysis appeared reasonable and thorough.
Proof of the effectiveness of L corrective action; I hcwever, will not be available until the next refueling.
(3) Valve Performance History 'I
The licensee has established a computer based valve performance , hi sto ry.
This program is a useful tool for evaluating needed , , improvement in valve performance.
., The inspector considered these actions as positive steps to establish - and maintain containment integrity.
Within the areas inspected, no violations or deviations were ! identified.
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Onsite Followup of written Reports of Non Routine Events at Power Reactor Facilities (92700) a.
(Closed) LER 325/88-025, Failure to meet Technical Specification 3.6.1.2b as Revealed Through Local Leak Rate Testing During the 1988/1989 Unit I refueling outage, local leak rate testing of primary containment isolation valves indentified nonquantifiable - leakage rates on eight valves. The inspector performed a detailed review of the local leak rate history for feed water inboard check valves B21-F010A and F010B during an inspection documented in NRC Inspection Report Number 50-325/89-32 and 50-324/89-32. During.that inspection, the inspector also examined the history on these same two Unit 2 valves and the LER was left open pending review of local leak rate test (LLRT) results for the Unit ' 2. feed water check ' valves during the current refueling outage.
The inspector examined the results of the LLRT performed on the Unit 2 valves in September 1989.
Both valves failed. The cause of the LLRT failures was attributed to wear of the valve seal - ring, i.e., the ethyl propylene (EPR) sof t seat material. The valves were repaired under work request 89-AUXVI and 89-AHDZ1 and retested in January - February 1990. In a letter to the NRC-cated November 27, 1989, Serial: NLS-89-312, Subject: Response to Request for Additional Information, LLRT Testing Frequency, the licensee committed.to modification of >the feedwater check valves B21-F010 A and B, which includes relocation from a region' of turbulant flow to a region of laminar flow or replace the valves with a new design or both of these actions, Although the engineering . study is not finali.md, management's current thinking is that both of j these actions are necessary.
The inspector also examined the lo' cal leak rate test history for the other six containment isolation valves which failed the' LLRT during i the 1988/1989 Unit 1 outage. These valves were_as follows: Outboard main steam isolation valves -(MSIV) B21-F028B and -- B21-F028D Reactor core isolation cooling (RCIC) pump discharge valve -- E51-F013 Residual heat removal (RHR) suppression pool cooling inboard -- valve E11-F024B Containment atmosphere control reactor building vacuum breaker -- CAC-X20A Reactor building closed cooling supply isolation valve RCC-V28 -- The LLRT history disclosed that valves B21-F028B and F028D, and CAC-X20A have also experienced repetative failures.
The licensee ' identified misal10nment of the valve disc seat to disc guide as the l %:.
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, cause of the leakage on valve B21-F028B.
The cause of the leakage for valves B21-028D and CAC-X20A is believed to have been caused by a foreign object being caught between the disc and seat during valve closure.
The cause of LLRT failures for the remaining three were as normal wear for valve E51-F013, valve body erosion for valve E11-F024B, and deposits on the seating surfaces for valve RCC-V28.
The inspector examined the maintenance work orders listed in table below which document repair of the valves.
TABLE - Maintenance Work Order Valve Number 88-BD SW1 B21-F028B 8S-BD SX1 'B21-F028D 88-BFDH1 CAC-X20A 88-BEMS1 E11-F024B 88-BCWII E51-F013 89-ABLK1 RCC-V28
The inspector also examined the oost repair LLRT data. - Review of this data showed.the_ repaired valves met LLRT acceptance criteria.
This LER is closed on the basis that cause of the LLRT failures have been identifed by the licensee and additional long term corrective actions required to prevent reoccurrence of the repetative LLRT failure are under active review by the Licensee.
NRC-will review LLRT results in future inspections, b.
(Closed) LER'324-88-002 and Supplement ! In a supplement to LER 324-88-002, dated August 22, 1988, the licensee reported the failure of the outer drywell head seal and feedwater valves B21-F010B (inboard check valve) and B21-F032B (outboard MOV) during the Unit 21988 refueling outage. The drywell head seal f ailure was thought to be due to a manufacturing flaw and was replaced with a new seal.
The failed seal in B21-F010B was replaced with a different seal considered to be of better quality and
the asbestos packing in B21-F032B was replaced with Chesterton-style ! 5300 packing.
These corrective actions were only partially effective. In the-'1989 Unit 2 refueling outage the drywell head outer seal, B21-F010B and ',. B21-F032B again failed the local leak rate tests performed in i September 1989.
The head seal and B21-F010B were the same type of ! f ailure experienced in 1988.
B21-F032B was a seat f ailure rather than a packing fai",ure as experienced in 1988.
In December 1989, the licensee received the - results of extensive laboratory analysis performed at the Harris E & E Center which ) I indicated that the head seal failure is a result of using Nickel ' Never-Seez lubricant in a high temperature, air enviornment.
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. licensee has taken steps to correct - this condition in that no lubricant was used in installation of the seals, and action was taken to improve insulation in.the drywell to reduce the temperature at'the-drywell head.
Engineering review on the feedwater valve problems is discussed in the previous paragraph.
This LER is closed on the basis that the problems are identified and the resolutions are under active review
by the licensee. The NRC will track the results of licensee action. through the routine leak rate test inspection program.
6.
Licensee Action on Previous Inspection Findings (92701, 92702) a.
(Closed) Inspector Followup Item (324, 325/88-11-01): Vent Path Model. The containment vent paths are capable of removing decay; heat' from the containment through hard piping vent lines.
However, the low pressure ductwork installed at transitions to the ' standby gas ~ treatment system and reactor building purge exhaust system fan would be overstressed and possibly fail in. accident conditions.
The licensee has evaluated this problem in response to NRC Generic Letter (GL) 89-10, Installation of a Hardened 1Wetwell Vent, issued on September 1, 1989. The licensee responded to GL 89-16 in a letter-dated October 27., 1989.
The licensee committed to modify the. vent system by 1993 or the second refueling outage from the date of their response, whichever is later.
The licensee is a member of a BWR Owners Group which.is currently developing generic ' design criteria '. for the hardened vent piping. This criteria, which will be available by April 30, 1990, will be utilized in any modifications to the vent path, b.
(Closed) Violation (324/89-32-02): Unauthorized Manipulation of a Valve.
The licensee's corrective actions for this violation are stated in
their December 8, 1989, response to NRC. The cause of this violation was attributed to an isolated error on the part of the ISI Group , technician who opened the tagged out valve approximately one turn.
The licensee issued nonconformance report (NCR) S-89-085 to investigate and correct this problem.
The licensee's corrective action included special counselling of the ISI group on this event, emphasizing the criticality of clearance tag violations.
The incident was also discussed with the Technical Support staff during'a training session. The inspector reviewed NCR S-89-085 and the training records documenting training on this event.
7.
Exit Interview l The inspection scope and results were summarized on February 20, 1990, ! with those persons indicated in paragraph 1.
The inspectors described the
areas inspected and discussed in detail the inspection results.
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. rate was acceptable at 0.308 wt. percent per day. No'significant problems.
were noted during the hydrostatic test; of the recirculation piping.: ,, , , ,,
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