ML20236L390

Provides Addl Info Re Performance Testing of Relief & Safety Valves,Per NUREG-0737,Item II.D.1 & 870617 Request. Evaluation of 83 Supports for Original Design Loads Performed.Mods Identified & Installed

ML20236L390
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	10/30/1987
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM TAC-44625, NUDOCS 8711100309
Download: ML20236L390 (7)

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. -m Postland GeneralElectricCortpony David W. Cockfield Vice President, Nuclear

' October 30, 1987

- Trojan , Nuclear ' Plant' l Docket 50-344 l License NPF-1

-U.S. Nuclear Regulatory Commission ATTN: Document Control Desk- j Washington DC 20555

Dear Sir:

. Performance Testing of Relief'and Safety Valves NUREG-0737. Item II.D.1: TAC No. 44625 By letter dated June 17 1987, the Nuclear Rigulatory Commission requested ,

that Fortland General Electric Company (PCE) provide additional information regarding performance testing of relief and safety.' values at Trojan.

Evaluation of the five items in the enclosure to'the June 17, 1986 letter has.been completed and PGE's responses are provided in the attachment to this letter. Special notice should be given to Items 3 and 5. 'These items involve the verification of' torque switch settings, and the development of

.a method to ensure' continued reliable operability-of-the safety' valves following a lift that involves discharge of loop seal water, respectively.

These items will require additional time to implement and compinte.' The i method and schedule to address these items is described in tho. attachment.

Sincerely, i

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Attachment c: Mr. John B. Martin Regionni Administrator, Region V U.S. Nuclear Regulatory Commission Mr.: Dave Yaden.. Director State of Oregon Department of T.nergy Mr. R. C. Bart NRC Resident Inspector Trojan Nuclear Plant h Q11$$o0$$ p P

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' 121 S.W Safrnon Street, Port:and Oregon 97204

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- Trojan Nuclear Plant' Document Control Desk Docket.50-344 October 30, 1987.

License NPF-1 Attachment Page 1 of 6 1

1 PORTLAND GENERAL ELECTRIC COMPANY (PGE) i RESPONSE TO NUCLEAR REGULATORY COMMISSION (NRC) l

REQUEST FOR ADDITIONAL INFORMATION - NUREG-0737. ITEM II.D.1 .

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1. NRC Request:

The.Impell report'on Trojan;provided piping support =1oads but did not j evaluate these loads. Also, the report provided the load combinations j analyzed, but did not provide the' allowable for each load combination.

a. Discuss how the piping supports were evaluated. l l

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b. Identify.the code used in the support analysis.

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E c. Provide the allowable stress for each piping support load combination.

d. Provide a table which compares the calculated and allowable stress for  ;

'the most' highly. loaded supports. Provide this table for supports both l upstream and. downstream of the valves. I i

PGE Response:

, a. The 83 uupports were evaluated for original design loads developed for L the NUREG-0737 analysis. Modifications were identified and installed.

The loads from the as-built piping' analysis were incorporated into the support calculations. The various parts that make up a pipe support were evaluated to their respective allowables as described in Item 1.c j below. '

b. Design codes used in pipe support evaluation:

+ American Society of Mechanical Engincors USA Standard (USAS) B31.7  !

1969 Ecition including addendas through 1971.

• Amerie:r. National Standards Institute (ANSI) B31.1 1973 Edition.

, + Manufacturers Standarolsation Society (MSS) -SP-58 1975 Edition.

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l + American Institute of Stet 1 Construction (AISC) 7th Edition.

+ American Society'of Mechsaical Engineers (ASME) -III Subsection NF J

(Various Editions) by vandor for standard components with load 'l capacity data sheets. l l c. Allowable stresses and loads for the design load combinations:

+ Ldad rated standard components - Allowabic loads for normal, upset, y emergency, and faulted loads as established by the component vendors j on their load capacity data sheets. -)

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4 Trojan Nuclear Plant Document Control Desk Docket 50-344' October 30, 1987 License NPF-1 Attachment Page 2 of 6

+ Structural steel - Designed to AISC with the following factors per the Final Safety Analysis Report:

Load combination Allowable Stress Normal AISC A Upset 1.25 x AISC Emergency 1.50 x AISC*

Faulted 1.50 x AISC*

Hydrotest 1.50 x AISC*

• Shear stress limited to 0.5 Sy

+ Steel for component structures - Designed to MSS-SP-58 with the following factors:

Load Allowable Stress Combination Tennion & Bending Shear Normal S 0.8 S Upset 1.2 S. 1.20 x 0.8 S Emergency 0.9 Sy 1.33 x 0.4 Sy Faulted 0.9 Sy 1.33 x 0.4 Sy Hydrotest 0.9 Sy 1.33 x 0.4 Sy

+ Welding of structural steel - Designed in accordance with AISC with the following factors:

Load

- Combination Allowable Stress Normal S = 18 kai Upset 1.25 S = 22.5 ksi Emergency 1.50 S = 27.0 ksi Faulted 1.'50 S = 27.0 kai Hydrotest 1.50 S = 27.0 ksi

+ Welding of MSS-SP-58 Components - Designed in accordance with ANSI B31.1 with the following factors:

Load 3 Combination Wold Stress Allowable Normal 0.75 S Upset 1.2 x 0.75 S Emergency 2.0 x 0.75 S Faulted 2.0 x 0.75 S Hydrotest 2.0 x 0.75 S

. l Trojan Nuclear Plant Document Control Desk Docket 50-344 October-30, 1987 License NPF-1 Attachment 1 Page 3 of'6

+ Expansion bolts.

Load Combination Allowable Loads - 1 in, diameter All Shear 5.5 k ,

Tension 4.1 k

d. Calculated loads and stresses versus allowables for supports with the least design margin:

1 Load Calc. Load / I Pipe Support Number Combination Component Stress Allowable RC-2501R-12-1-SS1047 Faulted 1/4 in. 1.08 k/in. 3.2 k/in'.

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Fillet RC-601R-2-1-SS302 Faulted 1/4'in. 2.31 k/in. 3.2 k/in.

! Fillet Faulted PSA-35 18.4 k 72.45 k l

RC-601R-2-1-SR312 Upset 1/4 in. 1.7 k/in. 1.9 k/in. I Fillet 'I l

RC-601R-2-1-H26 Hydro W4 x 13 fb = 9.13 ksi fb " 14 S k81 i RC-601R-2-1-SR304 Faulted -1 in diame- T = 3.5 k Ta = 4,1 k ter - Exp. V=0 Va = 5.5 k ,

Bolt )

f RC-2501R-12-1-SR15 Faulted Plate 15.5 kal 18 kai (shear)

NOTE: Supports on isometric RC-250lR-12-1 are upstream of the safety valves and supports on isometric RC-601R-2-1 are downstream of the safety valves.

2. NRC Request:

The PCE submittal stated that slug diversion devices (SDDs) were to be installed and support modifications were to be made at Trojan, but it did not state when these changes were to be made. State when the modifica .

tions were made or provide a schedule showing when they will be made. In its submittal PCE stated the as-built system would be evaluated to verify the Impell analyses based on the design data accurately reflect the' .

as-built system. If the system modifications have already been made, pro- 1 l

vide the results of this evaluation for our review. If the modifications have not boon made, the schedule requested above should include a date by which the results of the evaluation would be provided to the NRC.

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4 Trojan' Nuclear Plant - Document Control Desk Docket 50-344- October 30, 1987 License NPF-1 Attachment Page 4 of 6 PCE Response:

Four slug diversion devices were installed during the 1983 refueling out-age; one for both power-operated relief valves (PORV) and one for each of the three safety valves. As-built thermal-hydraulic and piping stress analyses were performed.for the modified system by Impell. The analysis results, as documented in Impell calculations, demonstrate satisfactory 'l qualification of the pressurizer safety and relief valvo piping system to j the applicable code (except as noted in Item 4 below). Support loads from I the Impell as-built analysis were forwarded to Bechtel and, as described in Item 1 above, used to confirm the structural adequacy.of the as-built pipe supports.

3. NRC Request:

l The Trojan Velan block valves use Limitorque SMB-000-25 operators while I the Velan block valves tested used Limitorque SB-000-15 and SMB-000-10 operators. Provide the torque switch setting and the torque produced by the plant operators at this setting. Verify that the torque produced by j the plant operator is 82 foot-lbs (the minimum torque tested to success-  !

fully close the valve). It is the staff position that proper operation of the valve cannot be concluded based solely on manufacturer's calcula-tions. The problems encountered with Westinghouse gate valves on closing, which were traced to the calculations used to size the operator torque requirements, indicate the need to experimentally verify the adequacy of the block valve / operator combination.  ;

1 PGE Response:

The torque switches for the Trojan Velan block valves are set at i 124 foot-lbs. The actual torque produced by the operators has not been i verified. The operator torque will be measured and verified'during the 1988 refueling outage as part of the Motor-operated Valve Analysis and  !

Test System (MOVATS) program. The torque switch setting of 124 foot-lbs l

should ensure the minimum of 82 foot-lbs necessary to close the valve. j

4. NRC Request:

The calculated piping load at one point upstream of the safety valve was said to slightly exceed the design load (see Page 5-2 of Impell h9 port 01-0300-1291, April 1984). The calculated load was also said to be ,

conservative. Provide an exact comparison of the calculated and allowable j stresses at this point.- Discuss the amount of conservatism in'the calcu- l lated stress. Justify not changing the system to bring it into confor-mance with Code standards.

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i Trojan Nuclear Plant Document Control Desk f.

l Docket 50-344 October 30, 1987 License NPF-1 Attachment ('

Page 5 of 6 PCE Response:

'The data for the point where the calculated piping load slightly exceeded design load is as follows, with stresses expressed in pounds por square inch: y i

Data Valve Combined Stress l Point Pressure OBE Grav Disch Eqn 9 Stress Allowable Ratio 84 11,465 13,818 195 115 25,445 24,120 1.05 The table indicates the calculated load exceeds the allowable by 5 per-cent. This is not considered significant due to conservatism in the analysis. Consideration of the following factors would reduce the cal-culated loads and eliminate the concern of higher-than-allowable stress at this point, j

a. By using thick wall theory to calculate pressure stresses, a substan- i tial reduction in calculated stresses would be shown for thin 6-inch Sch. 160 pipe. Thick wall theory would also show maximum pressures stresses to be on the inside surface of the pipe.

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b. Maximum stresses due to Operating Basis Earthquake (OBE) (seismic) are- ]

calculated to be at the outer surface of the pipe. Combining seismic '!

loads with pressure, and considering the variation in stress through the wall, would further reduce combined stresses.

c. The seismic analysis was performed using a 0.5 percent damping spectra for OBE response spectrum analysis for the piping in accordance with' the original methodology. Although this 0.5 percent damping valve is stated in Trojan's FSAR, it is considered very conservative. Use of higher damping spectra in accordance with standard industry practice would substantially reduce calculated seisntic stresses.

To accurately quantify the conservatism would require reunalysis of the piping system. It is pCE's judgement that suf ficient conservatism exists in the current analysis to bound tho 5-percent difference between the calculated and allowable stress.

5. NRC Rt. quest:

Two tests woro performed on the 6M6 valve with hot loop seals. These were tests 1415 and 1419 with loop seal temperatures of 290*F and 350*F, respectively. These tests raise a concern with respect to the Trojan 6M6 safety valves because the EPRI tests showed that the valves tested generally performed better with hot loop seals. These tests used ring settings (-77 -18) similar to the Trojan valves

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, .>.. l Tro'an j Nucicar Plabt s Document. Control Desk' Docket 50-344. October 30, 1987 License NPF-1 Attachment .q Page.6-of 6 I

In Test 1419, the : valve' reopened. af ter closure and chattered,-- and the ' test was terminated after the valve was manually opened to stop.the chattering,

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This-result does not indicate a valve closing problem for the Trojan . . i safety valve since an identical test (Test,1415).had already demonstrated' that the valve performed satisfactorily and' exhibited no sign of instabila ity. The closing chatter in Testf1419 may.possibly be a result of the.

repeated actuation of.the-valve in loop seal and water discharge tests.

The 6M6 ' test valve was subjected to 17 steam, water. . and transition tests. In the first four or five tests the valve fluttered and chattered- _j during loop seal discharge but stabilized and closed-successfully,- After' Test 913, there were four; instances in which the test was terminated due to chattering on closing. -Called guiding surfaces and damaged parts-were refurbished or replaced before the next test started. .The test results j showed that the valve performed well after each repair,-but the closing.

chatter recurred in the subsequent test. Test 1415 was performed immedi-~ ,

ately after valve maintenance and the valve performed stably. The next ;t test (Test 1419) encountered chatter in closing even though it was'a.

repeat of Test 1415 at similar fluid conditions. .This' suggests'thati inspection and maintenance _are important to the. continued operability of this valve. Therefore, the licenseo must develop a method to ensure con-tinued reliable operability of the safety. valves following any lift of the valves that involves the discharge of~1oop seal water (eg, inspection and maintenance procedures shall be developed and incorporated into the Plant-operating procedures-or licensing documents such as the Plant Technical' specifications). .s

'i PCE Response:

Item 5 requires the development of procedures or other methods to ensure operability of the safety valve following a lif t and subsequent' loop seal water discharge. The Trojan Plant. Operating' Manual will be revised to incorporate valve inspection and maintenance requirements ,1f a lif t of 'the safety valves occurs. Plant procedures will require inspection and main-tenance of the lifted valve (s) to be performed prior to resuming power operation. The procedures will be revised by the end of the 1988 refuel-ing outage.

DRS/KLB/kal 1854P.1087

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