IR 05000413/1990021

Insp Repts 50-413/90-21 & 50-414/90-21 on 900627-29.No Violations or Deviations Noted.Major Areas Inspected:Main Steam Safety Relief Valve Setpoint Testing & Maint & Repair of Coatings on Steel Containment Vessel

ML20055H421
Person / Time
Site:	Catawba
Issue date:	07/19/1990
From:	Belisle G, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20055H420	List: IR 05000413/1990021 ML20055H419
References
50-413-90-21, 50-414-90-21, NUDOCS 9007260179
Download: ML20055H421 (6)

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f aO . , , UNITED STATES ~

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ReportLNos.': 50-413/90-21 and 50-414/90-21

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Licensee': Duke Power Company

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422 South Church Street i Charlotte, NC 28242  ;

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- D'ocket Nos.

50-413 and 50-414- License Nos.: NPF-35 and NPF-52-

- Facility Name: Cctawba' Units 1 and-2

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Inspection Conduct d: Ju .27-29, 1990 i

Inspector:- , / o .

J- J/AX ahan

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'6 proved by:_ l }/ M M y N

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'G.W. Bert1rle, Chief

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Test Programs Section ,

3 Engineering Branch y Division of Reactor Safety J

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SUMMARY

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- Scope: 'd -

s-This routine unannounced l inspection'was conducted .in the ' areas of main steam I

. safety relief valve ' set point' testing - Unit 2, maintenance' and- repair of coatings on the steel' containment vessel'- Units 1'and 2,.and' action on previous R, i inspection finding .

. Results:-

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?q - The licensee's approach to resolving problems identified during testing of main ' d

~;, ' steam -safety relief valves was conservative,- timely, and thoroug ,

.e In the areas inspected, violations or deviations were not identified,

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.13 REPORT DETAILS- Persons Contacted ,

Licensee Employees }

K. Bishop, Mechanical Engineer, Maintenance Engineering

• C. Cauther, Mechanical Engineer, Maintenance Engineering

• J. Glann,-Compliance Engineer .

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H. Nekooasl, Civil Engineer, Maintenance Engineering

'*T. Owen, Station Manager

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NRC Resident. Inspectors

• Orders, Senior Resident Inspector J. Ze11er, Resident Inspector

• Attended Exit Interview Main Steam Safety Relief Valve Set Point Testing - Unit 2(61701)

On January 26, 1990, while the licensee was che: king Unit 1 safety relief valves (SRVs), the valve manufacturer's repretentative . informed the

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licensee that the constant associated with the ; ydrcset Test Device had recently been changed from a value of .344 to a visue of .352. _ This change was a result of testing performed by- the manufacturer in August 1989, but the licensee was not informed. The licensee adjusted the Unit 1 SRVs using the revised- constant. The licensee issued Problem

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Investigation: Report (PIR) number 0-C90-0026 to evaluate the. operability of the. Unit 2 SRVs since Unit 2 was operating. The-licensee recalculated the as-left setpoints for the Unit 2 SRVs using the new constant value, and determined that the valves were set to lift at values slightly above-th'r plus 1 percent tolerance permitted by Technical Specification (TS)

Tible 2.7-2. rHowever, the licensee concluded that the valves would perform their function, and the slight increase in the. lift setpoints had no: safety significance. On February 6.1990, the licensee requested a s Waiver Jof Compliance. from the NRC pertaining to the failure of the SRV setpoints to meet .the 1 percent TS tolerance. The Waiver of Compliance stated that the revised SRV setpoints were 1.5 percent above the TS Table 3.7-2~ values for a few' of the Unit 2 valves, while the others met the 1 percent tolerance.- The Waiver of Compliance was granted by the NRC in a letter dated February 6, 1990, SUBJECT: Temporary Waiver of Compliance -

Catawba Nuclear Station, Units 1 and ' 2 (TACS 75884/75885), which was subsequently issued-as Amendment No. 64 to the Technical Specification t

'The inspector examined results of setpoint testing performed on the Unit 2 SRVs in May 1990. Technical Specification 4.0.5 requires that SRVs be ,

. tested in accordance with Section XI of the ASME Code during' tach refueling outage.Section XI requires that a minimum of 20 percent of the valves be tested. Additional valves were required to be tested if any in i

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the- original sample did not meet the setpoint requirements. Due to the '

change in the Hydroset Constant. the . licensee was required to test and '

adjust' all 20 Unit 2;SRV The inspectcr reviewed procedure MP/0/A/7150/72,- Main Steam Safety Valve Setpoint Test.- This' procedur !

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specifies requirements for pe f'rming the tests, including prerequisites,.

test methodology, and acceptance criteria. .The inspector examined the

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test data and verified that the revised Hydroset Constant Value, 0.352, was used in the SRV setpoint calculations. The following is a summary of L the SRV setpoint results:

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. Required * Adjusted l Valve As-Found Lift (PSIG)

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Number Value (PSIG) -Setting (PSIG) __ Value 2-SV-2 1208 1175 1178- ,

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2-SV-3 1225 1190 1187 2-SV-4 1240 1205 1208 2-SV-5 1240 1220 1213 2-SV-6 1250 1230 1232 >

2-SV-8 1210 1175 1179 2-SV-9 1210 1190 1187 2-SV-10 1222 1205 1208

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2-SV-11 1235 1220 - 1220 2-SV-12 1230 1233

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1241 2-SV-14 1200 1175 1173 2-SV-15 1225 1190 1198 2-SV-16 1230 1205 1198 2-SV-17 -1250 1220 1215 2-SV-18 1250 1230 1222 2-SV-20 1195 1175 1180 2-SV-21 1210 1190 1193-

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2-SV-22- 1244 1205: 1198

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2-SV-23 1238 1220 .1213- t 2-SV-24 1245 1230 1236

• The . required set pressute had a tolerance of plus or minus _ percent until the first outage after February 6,1990, the date of approval of Amendment No. 64 to TS Table 3.7-2. The tolerance i is _ now:specified to be plus or minus one percent in accordance ~with standard TS requirement The valve manufacturer, Dresser Industries, Inc., submitted a formal notification to-the licensee in a letter dated May 17, 1990, SUBJECT: 10 i

- CFR Part 21 Notification,1566 Hydroset New Value Constant K Factor, Duke Power's Request to NRC Relating to.New Constants, regarding the August <

i , 1989' refinements to the K factors. These revisions were made as-a -result l' of additionea testing performed on various size valves. The change in the p

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constant values ranged'from 0.2 to 4.0' percent.

j Within the areas inspected, no violations or deviations were identifie ,

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3. - Maintenance and Repair of Coatings on the Steel Containment Vessel - Units-1.and2(62700).

Corrosion 'of the McGuire Unit 2 steel containment vessel. which was discovered by the licensee prior-to performing an integrated leak rate test, prompted an inspection of the Catawba steel containment vessel for l similar type degradation. The Catawba inspection revealed a small amount-of corrosion on the vessel and some~ coating failures. The cause of the corrosion 'was attributed to standing water in the annulus area. The source of: the water. was leaking instrument lines. The water.came in !

contact with the steel containment vessel because the slope of the annulus floor was not sufficient to direct the leakage to the installed floor

~ drains. Small pin hole imperfections in the protective coatings permitted the water to penetrate the coatings and caused corrosion of the steel containment vessel and additional- coating = failure The problem was not easily detected because of heating, ventilation, and air conditioning ducts installed on the annulus floor adjacent to-the steel containment vessel.. which made those areas inaccessible for inspectio The licensee's corrective actions included removal of the duct work, removal of the failed coatings and any: corrosion, a'd recoating the steel vessel- l with a coating material qualified for servico in immersion condition The inspector walked down the Unit 2 annulus whom work was in progrese to remove . the damaged coatings and the duct . work. This work was approximately half completed. The inspector examin?d the steel v.ssel

- and noted that the base metal had not been affected by the coating failure The corrosion of the steel was minor or !nsignifican The licensee has finished the repairs .in the Unit 1 annulus are The inspector reviewed the following ' quarterly records documenting the Unit 1 .

repairs: '

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Work Request Number'3510 NSM for removal of duct. work and support Work Request . Number 3511 NSM for removal of concrete anchors from duct work' supports and repairs to concrete floor where anchors were remove .

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Work Request Number 1717 MES which was- the test criteria to verify that the duct' work could be removed without affecting the negative pressure distribution in-t.1e annulus are p ,1: - Work Request Number 1719 MES which covered removal of the old ;'

ccatings:and replacement of the coating with materials meeting Design

,1 Specification 303, Nuclear Coating Maintenance, Coating of Carbon Stee Records documenting results of quality control inspections performed on the required coating Post Maintenance Testing which was performed using procedure PT/1/A/4450/03C, Annulus Ventilation System. Performance Tes The

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test . verified that after modifications to the duct work were

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completed, negative pressure distribution in the annulus still met design requirement '

Within the areas inspected, violations or deviations were not' identifie ., Action on Previous Inspection Findings (92701)

, (Closed) Unresolved ' Item (413/88-04-01): Requirements for- Functional

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Testing of Snubbers '

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During' functional testing .of Unit 1 snubbers in October 1989, 3 of the >

initial. sample of 37-snubbers failed to meet functional: test requirement i The' license then selected an additional random sample of 54 snubbers'fo functional testing per the requirements of TS f 4.7.8 e.2 and TS Figure

4.7-1. Five of these snubbers failed the functional test. The licensee selected an additional random sample of 36 snubbers. Two additional snubbers, both size PSA 1/2, failed the functional test in this grou ,

Since seven of the. total ten functional test . failures were size PSA :1/2 snubbers, the licensee designated these failures as a failure mode group ;

in accordance with information contained in a Draft of Revision 2 of ANSI OM-4, . Examination and Testing of Dynamic Restraints (Snubbers). This method is documented in a Technical Specification. Interpretation prepared by.the licensee in accordance with 0M-4 and in discussions with personnel from the Office of~ Nuclear Reactor Regulation (NRR) on November 10, 198 However,. by use of this TS Interpretation, the licensee did not test enough snubbers larger than size PSA 1/2 to meet the requirements of TS 'i Figure 4.7-1- for the three other failures .(TS Figure 4.7-1 requires

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testing 91 of these other size snubbers, wherein the licensee only tested "

71). Therefore, the inspector concluded that the licensee TS Interpretation

. appeared 'tc conflict with the current:TS. . The licensee submitted a TS 4_ amendment. request to NRR on April- 15, 1988, to formally incorporate this interpretation into TS 3/4.7.8, In a letter dated September 26, 1988, NRR informed the licensee that this TS -amendment was .still under review. . In -

an . inspection conducted February 28 to March 3,,1989 (documented in NRC Inspection Report. Nos. 50-413/89-06 and 50-414/89-06), the inspector ,

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.discu;!e. the use of this TS Interpretation with the license Subsequent J to the irepection, additional telephone discussions were. held between Regior 11, NRR, and licensee personnel regarding the use of this TS f , interpretatio As a result of these discussions, the licensee canceled the TS ' interpretation since it appeared to conflict.. in part, with the

, < current approved TS. . The inspector reviewed the results .of snubber-3- functional testing during the two subsequent f . refueling outages (performed on Unit 1 snubbersEnd.of Unit 1, Cycles 3 and 4) an

_ during the last refueling outage (End of Unit 2, Cycle' 2). This review

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r disclosed that none of the 74 Unit 1 snubbers selected for testing (2 1 ~ random samples of 37) failed the functional test; however, 1 PSA mechanical snubber failed. the functional test during the Unit 2, End of Cycle 2 ' testing. Some functional test failures were experienced while

,. testing' Anchor Darling mechanical snubbers. However, they were documented

, .on PIR C-89-150. The inspector reviewed the PIR and concluded that'the licensee had ' properly evaluated and resolved problems regarding the Anchor

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Darling, mechanical snubber functional test- failures. Based on the additional snubber functional testing performed since this Unresolved item was . originally identified. the inspector concluded ' that the- apparent failure to test an additional 20 PSA snubbers larger than size PSA 1/2 during., the Erd- of Cycle 2 outage had no safety significance. The

. licensee's snubber functional testing program. meets current TS require-ment , Exit Interview The inspection scope and results were summarized on-June 29, 1990, 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 report. Dissenting comments were not received' from the license ,

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