IR 05000424/1989012

Insp Repts 50-424/89-12 & 50-425/89-14 on 890313-17.No Violations or Deviations Noted.Major Areas Inspected:Snubber Surveillance Program,Unit 2 Containment Structural Integrity Test Program & Piping Vibration Testing Program

ML20245G479
Person / Time
Site:	Vogtle
Issue date:	04/19/1989
From:	Belisle G, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20245G445	List: IR 05000424/1989012 ML20245G441
References
50-424-89-12, 50-425-89-14, NUDOCS 8905030156
Download: ML20245G479 (8)

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,f ug% NUCLEAR REGULATORY COMMisslON

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Report Nos.: 50-424/89-12 and 50-425/89-14 Licensee: Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 Docket Nos.: 50-424 and 50-425 License Nos.: NPF-68 and NPF-81 Facility Name: Vogtle 1 and 2 Inspection Conducted: March 13-17,1989 lospector:  ! h///9 Date Signed J.J.Legah&h j Approved by: hC&// '@

G. A. Bel i sl e , Chie'f M///[N Da te ' Signed Test Programs Section Engineering Branch Division of Reactor Safety ,

SUMMARY Scope: This routine, unannounced inspection was in the areas of the Unit I snubber surveillance program, the Unit 2 containment structural integrity test program, the Unit 2 piping vibration testing program, the Unit 2 thermal expansion test program, reactor coolant system leakage, and followup on actions of previous inspection finding Results: Within the reas inspected, violations or deviations were not identifie Piping thermal expansion and vibration test results indicate that the licensee is performing these tests in accordance with NRC require-ments. The licensee's methods for ccrrecting test deficiencies were conservative. Licensee management is involved in the test results review proces The licensee has implemented a snubber functional testing program which gaes Svond minimum NRC requirements. The licensee has taken a conservative ap,aroach in interpreting Technical Specification- (TS)

snubber seveilici. e requirement g g $ ,8 4 u

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i REPORT DETAILS Persons Contacted Licensee Employees

• G. Bockhold, General Manager

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J. Davis, Maintenance Engineer

• C. Gabbard, Senior Regulatory Specialist- l M. Hickcock, Snubber Engineer )

H. Handfinger, Maintenance Superintendent  !

• J. Williams, Plant Engineering Supervisor  !

D. Zivkovic, TG&V Test Supervisor j i

Other licensee employees contacted during this inspection included two engineers, three mechanics, and administrative personne ,

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-NRC Resident Inspectors J. Rogge, Senior Resident Inspector (Operations)

C. Burger, Senior Resident Inspector (Construction)

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• Attended exit interview Snubber Surveillance Program - Unit.1 (70370)

The inspector re/iewed procedures and quality records related to the Unit I snubber surveillance program. Acceptance criteria examined by the inspector appear in Unit 1 Technical Specification 3/4. ' Snubber Surveillance Procedures Review The inspector examined the following procedures which control the snubber surveillance program:

85056-C, Visual Examination of Snubber C, Functional Testing of Mechanical and Hydraulic Snubber C, Wyle 100 Snubber Test Machine, Quality Records Review The inspector reviewed quality records documenting results of functional testing performed on Unit 1 mechanical snubber during th October-December 1988 refueling outag These records included a summary of the functional testing results and data for tests performed on various Pacific Scientific (PSA) and Anchor Darling (AD)'

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snubbers, under Maintenance Work Orders 18803802, 18807703, 18807967, i 18808029 through 18808031, and 18808036. Of the 547-snubbers tested, a 41 failed to meet the functional test criteria. A breakdown of the

, test failures per snubber type is as follows:

(1) Type I included all PSA mechanical ' snubber sizes. In the initial ~ sample of 55. snubbers' tested, . 3. snubbers failed the

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functional tes Per TS1 requirements, the licensee tested 28 -

additional. snubbers for each test failure. When the 84 (3 x 28); )

snubbers were tested, additional test failures were identifie q The licensee then performed additional testing and evaluation o ;

Based on the engineering .

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the functional test failure evaluation . results, the licensee determined that all' but one failure could be attributed to one of three distinct failure mode group These groups are summarized in the following table:

Failure Mode Group (FMG) Description Number 1 PSA size 1/4 and 1/2 169 )

snubbers  :

2 Snubbers on RHR system 13 l subject to water hammer f i

3 PSA size 10 snubbers 64- :

All snubbers in the FMGs were tested. In addition,'the licensee I tested an additional 115 snubbers from the remaining ~ size snubbers (PSA 1, 3, 35, and 100). A total of 38 of the 416 PSA snubbers failed to meet the functional test acceptance criteri All but one functional test failure occurred. in the snubbers grouped within the FMG (2) Type II included AD size AD 40 through AD 500 snubber From the initial 55 AD snubbers tested, 3 snubbers ' failed to meet functional test criteri Two of these snubbers had been installed on the portion of the RHR system subjected to a water hanne The licensee tested the remaining 3 AD snubbers '

installed on this portion of .the RHR system, 28 additional AD .,

snubbers for the one remaining failure group, and 28 more due to the RHR failure mode group for a total of 116 snubbers. No

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additional functional test failures occurre !

(3) Type III included AD size AD-1600 through AD 12500 snubbers.

l The test plan required testing ten percent of these snubbers.

l All 13 snubbers selected for functional testing passe !

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The inspector concluded that the licensee's snubber functional '

testing program was conservative and complied with TS requirement Within the area intpected, violations or deviations were not identifie . Containment Str'uctural Integrity Test - Unit 2 (63050)

The inspector examined a Bechtel Report titled, "Frimary Reactor Contain-ment Structural Integrity Test - Vogtle Unit 2, Final Report", dated December 1988. Acceptance criteria utilized by the inspector appears in l Final Safety Analysis Report (FSAR), Sections 1.9.18, 3.8.1.7.1, and !

14.2.8.1.100 and Regulatory Guide 1.18, Structural Acceptance Test for

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Concrete Primary Reactor Containments. The report summarized the test results and contained the test data and containment displacement versus pressure graphs for each extensometer location. All measured deflections were within acceptance test criteria. The containment test pressure was 1.15 times the design pressure per the guidance in RG 1.18, i.e. , (1.15)

(52 psig) = 60.2 psi The predicted versus measured deflections of the Unit 2 containment structure are summarized in the following table:

Predicted Displacement (Acceptance) Measured Locations Deflection Deflection Vertical Movement at Dome Apex 0.44 i .293 i Radical Movement at Cylinder 0.31 i .156 i Midheight Radical Movement at Equipment 0.56 i .269 i Opening Residual Displacement of Dome Apex 0.08 i .003 i Residual Displacement at Cylinder 0.06 i .012 i Midheight l The report evaluated the effect of the one extensometer which failed l during the test and its effect on other test data. No measurable cracks were identified during the Structural Integrity Test (SIT). The ::T was completed in accordance with preoperational test procedure 2-300-5, Containment Structural Integrity Test. The inspector witnessed the actual j test during a previous inspection and documented the results in NRC a Inspection Report No. 50-425/88-54.

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Within the area inspected, no violations or deviations were identified, i

4. Piping Vibration Testing - Unit 2 (70331 and 70531)

The inspector examined piping vibration test procedures. Acceptance !

criteria utilized by the inspector appear in FSAR Sections 3.9.8.2.1, l 14.2.8.1.14, 14.2.8.1.104, 14.2.8.1.114, and 14.2.8.2.43, and FSAR

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Questions Q210.40 and Q210.41. Procedures examined were as follows:

Preoperational test procedure 2-300-09, Power Conversion and Emergency Core Cooling System Dynamic Tes l l

Preoperational test procedure 2-3L0-11, Steady State Vibration Monitoring of Safety-Related Pipin l Startup Test Procedure 2-600-06, Dynamic Response Test i The preoperational test procedures provided instructions for verifying the .

acceptability of the piping response to steady state and transient '

vibrations during the preoperational testing which was conducted prior to fuel load. The startup test procedure provided instructions for verifying that selected safety-related and balance of plant piping systems responded in accordance with design under specific steady state and transient conditions during precritical heatup and power ascensio During the procedure review, the inspector verified that test prerequisites and acceptance criteria were specified and that test instructions and objectives were clearly state The inspector also examined quality records documenting the steady state vibration test results that were completed during the preoperational testing phase. These records included the completed and signed off procedure 2-300-11, the test log, test data, test exception reports, and test trouble sheets (TTS). The TTS documented problems identified during the vibration tests and the disposition and correction of these problem The records reviewed indicated that the test was completed in accordance with test instructions, FSAR Commitments, and NRC requirement Within the area examined, deviations or violations were not identifie . Piping Thermal Expansion Test (70370)

The inspector examined the thermal expansion test procedures, examined l

portions of safety-related piping systems that , e monitored during the thermal expansion test, and reviewed test dat Acceptance criteria utilized by the inspector appear in FSAR Sections 3.9.8.2.1, 14.2.8.1.103, and 14.2.8.2.48 and FSAR Question Q210.4 _ __-_-_

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a. Review of Thermal Expansion Test Procedures l

The inspector examined preoperational test procedure 2-300-08, j Thermal Expansion Testing, and startup test procedure 2-600-11, 8 Thermal Expansion Tes Procedure 2-300-08 covered testing of piping, prior to fuel load which has normal operating temperatures in excess of 250 F. This test verified that thermal movement was within j design limits. This test was also used to obtain data to size shims j required for primary equipment supports. Procedure 2-600-11 covers thermal expansion testing of piping during precritical heatup and ,

power ascensio During this test, the licensee will verify that; l safety-related and balance of plant piping can expand without being restricted, that the expansion is in accordance with predicted design !

limits, that shims for equipment supports and pipe whip restraints were sized based on measurements made during the preoperational thermal expansion test, that shims were properly fabricated and installed, that problems (restrictions) which were identified during the preoperational thermal expansion test were corrected, and that data for evaluating the surge line thermal stratification required by IE Bulletin 88-11 will be obtained. The inspector verified that test prerequisites and acceptance criteria were specified in procedures, and that test instructions and objectives were clearly state The !

inspector also examined Bechtal specification X4AQ588, Thermal i Stratification Data Collection Plan for Pressurizer Surge Line During Power Ascension Testing. This specification specifies the type and location of instrumentation and data required for evaluating the surge line thermal stratification issue identified in IE Bulletin 88-1 ;

b. Walkdown of Safety-Related Piping Systems Prior to this inspection, following fuel loading, precritical reactor ,

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coolant system (RCS) heatup was in progress. On March 10, 1989, while the RCS was at a temperature of 450 F, the licensee tested the RCS isolation valves on the safety injection lines in accordance with TS 4.4.6.2.2. One of the isolation swing check valves was found to exceed the maximum allowable leakage permitted by TS Table 3.4- When this occurred, the licensee decided to return the RCS to ambient temperature and partially drain the system in order to investigate the cause of the leaking valve and to correct the proble Resolution of this problems is further discussed in paragraph 6, bel o During the heatup to 450 F, the licensee identified two AD snubbers which were locked up. These snubbers 'were installed on hanger number V2-1305-064-H006, on the feedwater system, and hanger number V2-1302-107-H011, on the auxiliary feedwater system. These problems were documented on TTS 12 through 15 hanger l

V2-1305-064-H006) and TTS 16 (hanger V2-1302-107-H011)(. The licensee l disassembled the snubbers in order to determine the possible cause l for the snubbers locking up. The inspector observed disassembly of the AD model 12500 snubber from hanger V1-1305-064-H00 No deficiencies were observed during disassembly and the reason for this

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i snubber failure (locking up) was unknow The licensee also disassembled the model AD 503 snubber from hanger V2-1302-107-P01 When this snubber was disassembled, the pinion gear was found to be broken and the shaft was bent. This caused the snubber to lock u The licensee attributed the probable cause of the damage to this snubber as a possible overload due to water hammer. The inspector walked down portions of the piping systems in loops 1-4 and examined the feedwater, auxiliary feedwater, pressurizer surge, and RCS piping. During the walkdown, the RCS was at ambient temperatur The inspector verified that new snubbers were installed on the above hangers, and examined instrumentation installed on the pressurizer surge line to monitor temperature stratification. The inspector also verified that permanent plant equipment and temporary structures were not restricting pipe movement. The inspector also examined selected spring cans and verified that travel stops were remove Quality Records Review The inspector examined the data collected required by procedure 2-600-11 through the 450 temperature platea The inspector reviewed the test log, and TTS 001 through 003, and 005 through 01 The inspector also examined completed preoperational test procedure 2-300-08 and verified that the test had been signed off and completed in accordance with the test instructions specified by the procedur The inspector reviewed the test log and selected TTS. The inspector also examined snubber pre-service inspection record Within the areas inspected, no violations or deviations were identifie . Reactor Coolant System Leakage - Units 1 and 2 (70449)

As discussed in paragraph 5.b. above, a 10-inch diameter swing check isolation valve on the safety injection system was found to be leaking during routine TS surveillance testing. The leakagc was estimated to be 25 gpm versus the TS limit of 5.0 gpm. The licensee drained the RCS and disassembled the leaking valve (number 1204-46-086). Upon inspection, the licensee discovered that the disc arm was damaged at the point where it was engaged by the locking pin. The cause of the damage was unknown but was attributed to a possible water hammer. The valve was manufactured by Westinghouse. The licensee inspected the valve body and verified that it~

was undamage They then installed new valve internals (disc, disc arm and locking, and pivot pin assembly). The licensee disassembled the remaining three valves (numbers 7204-46-83, 84 and 85), inspected the valve bodies, and installed new valve internal No damage was noted by the :icensee during examination of these valve internal The inspector examined the valve internals from these three valves and concurred with the licensee. The valve internals were sent to Westinghouse for their examinatio In a letter from Westinghouse to Georgia Power Company dated March 16,1989, Westinghouse stated that, based on a review of quality assurance records, the materials used in manufacturing the valves complied with design requirement The vendor concluded that the damaged valve was J

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an isolated case based on the fact that the same valves have performed satisfactorily in other Westinghouse plants. The vendor also recommended that one valve be disassembled and visually inspected during the first refueling outage. During a meeting held in the Region II office on March 22,1989, NRC requested that the licensee perform material testing on the damaged valve to veri fy that the materials meet design requirements. The licensee agreed to perform this testin The inspector examined Operations Test Procedure 14450-01, RCS Pressure Isolation Valve Test. This procedure contains requirements for leak testing isolation valves per TS 4.4.6.2.2. The inspector examined a copy of this procedure which documented testing performed on January 27, 1989, on these same isolation valves which are installed in Unit The observed leakage was negligibl Within the area inspected, no violations or deviations were identifie . Action of Previous Inspection Findings (! '/01)

(Closed) Inspector Follow-up Item 424/86-53-01: Stressing Ram Calibration and Conversion Accurac The inspector examined the stressing ram calibration data performed on the stressing rams used to post-tension Units 1 and 2 containment tendons. The rams were calibrated prior to stressing Unit 1 tendons and again prior to stressing Unit 2 tendons. The inspector reviewed the calibration data and noted that there was good correlation between the two data set . Exit Interview The inspection scope and results were summarized on March 17, 1989, with those persons indicated in paragraph 1. The inspector described the areas inspected and discussed in detail the inspection results. Proprietary information is not contained in this repor Dissenting comments were not received from the licensee.