ML20237F905
| ML20237F905 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/06/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8708130223 | |
| Download: ML20237F905 (15) | |
Text
{{#Wiki_filter:s e.% L., -DUKE POWER GOMPANY P.<O. Hox 33180 - CHAHLOrn5, N.O. 28242 HALB. TUCKER Tzternown vmm encemen (704) o7& 4531' WUDLEAR PW4HHJCTION . August 6, 1987 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
Subject:
Oconee Nuclear Station i Docket'Nos. 50-269, -270, -287 " Performance Testing of. Relief and Safety Valves" Request for Additional Information
Dear Sir:
By letter dated April 22, 1987, the NRC requested additional information regarding NUREG-0737 Item II.D.1, " performance Testing of Relief and Safety Valves", for Oconee Nuclear Station. By a letter dated June 10, 1987, Duke Power ' Company (Duke) had advised the NRC Staff of a delay in providing a response. Attached, please find a response to requests 2, 3, and 4. Duke currently anticipates in providing a response to the remaining request for additional information by January 15, 1988. Very truly yours, 'd3c m Hal B. Tucker PFG/58/sbn Attachment, xc: Dr. J. Nelson Grace,. Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta,. Georgia 30323 Ms. Helen Pastis Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 i Mr. J. C. Bryant NRC Resident Inspector j gh Oconee Nuclear Stclion hO 23 B70006 E P CM 05000269 PDR
r__-. i l DUKE POWER COMPANY OCONEE NUCLEAR STATION ATTACHMENT NUREG-G737 ITEM II.D.1 " Performance Testing Of Relief And Safety Valves" Response To Request For Additional Information l
Request 2 In response to Question 3 Duke Power stated that low pressure PORV operability testing at each plant prior to start-up and low pressure testing at Wyle Labs demonstrate PORV operability at low steam pressures. Provide data or documentation from this testing to the NRC for review to support this statement. Response 2 is a copy of the Oconee Periodic Test Procedure "PORV Operability. Test (PT/0/A/201/04)". This test is conducted at less than or equal to forty-five (45) psig of steam prior to each unit startup from cold shutdown. is a copy of a typical Wyle Laboratories' Test Report. Valve actuation at low steam pressures (in this case forty-five (45) psig) is required to discharge any buildup to condensation that forms in the steam supply lines during heatup. The valve is alto actuated at low steam pressures to realign the main and pilot disc when seat leakage is encountered. This is normally done after rebuilding the valve (see Enclosure 3). Testing at Wyle Laboratory is performed each refueling outage. Request 3 In response to Question 10, Duke stated the PORV closing pressure was 2415 psia. This is higher than the maximum closing pressure of 2360 psia in the EPRI tests. To demonstrate the ability of the PORV to close at the plant setpoint, reference was made to tests at Wyle Labs with a valve upstream pressure of 2415 psia at closure. Provide test data or documentation from these tests to the NRC for-review to support the conclusion by Duke that acceptable valve operability has been demonstrated. Response 3 A copy of the Wyle Laboratories' Test Report documenting the elevated pressure test on an Oconee PORV, serial number BL 08903 is provided by Enclosure 2 of, Response 2. Request 4 Your response to our request for information stated the PORV control circuit components are located in the cable room and are not subjected to any harsh l environment and are fully qualified for that area. However, there must be some equipment, if only cables, connecting the components in the cable room to the PORV. In order to demonstrate the Oconee 1, 2, and 3 control circuitry is ( qualified, provide documentation to show the equipment has been qualified under 10 CFR 50.49, or, to allow a complete review of the qualification of the control circuitry for the PORV under NUREG-0737, provide the following: (A) Provide a list of all PORV control circuitry needed to mitigate NUREG-0737 transients such as the following: l I
l ] ] (1) Switchgear (2) Motor control centers (3) Valve operators and W i.enoid valves (4) Motors (5) Logic equipment (6) Cable (7) Connectors (8) Sensors (pressure, pressure differential, temperature, flow and level, neutron, and other radiation) (9) Limit switches (10) Heaters (11) Fans (12) Control boards (13) Instrument racks and panels (14) Electric penetrations (15) Splices (16) Terminal blocks (B) For each item of equipment identified in A, provide the following: (1) Type (functional designation) (2) Manufacturer (3) Manufacturer's type numb?.r and model number (4) Plant ID/ tag number and location (C) For each item of equipment listed in above, provide the environmental envelope, as a function of tima, that includes all extreme parameters, both maximum and minimum values, expected to occur during NUREG-0737 transients, including postaccident conditions. (D) For each item of equipment identified above, state the actual qualification envelope simulated during testing (defining the duration of the environment and the margin in excess of the design requirements). If any method other than type testing was used for qualification, identify the method and define the equivalent " qualification envelope" so derived. (E) Provide a summary of test results that demonstrates the adequacy of the qualification program. If any analysis is used for qualification, justification of all analysis assumptions must be provided. (F) Identify the qualification documents that contain detailed supporting information, including test data, for items D and E. Response 4 This request indicates that the NRC Staff believes that Item II.D.1 cf NUREC-0737 requires that the PORV control circuitry be environmentally qualified, pursuant to 10 CFR 50.49. Please be advised that Duke does not agree with this current l Staff understanding of the requirements specified by Item II.D.1. Duke's understanding is that Item II.D.1 does not require that the PORV control I circuitry be environmentally qualified. Notwithstanding a response regarding the environmental qualification of the PORV and the PORV control circuitry is l provided. l l l l i
. The PORVs at Oconee are non-safety and are not required to perform any accident mf.tigating function for the Ooncee Design Basis Accidents (DBAs). As such, the-PORVs and their associated circuitry are not within the scope of 10 CFR 50.49. 10 CFR 50.49 paragraph (b)(2) requires that non-safety equipment whose failure could prevent satisfactory. accomplishment of safety functions of safety-related equipment must be qualified. Those safety functions at defined in paragraph (b)(1) of 10 CFR 50.49 are as follows: (i) ensure the integrity of the reactor coolant pressure boundary, (ii) ensure the capability to shut down the reactor and maintain it in a safe shutdown condition, and j (iii) ensure the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to 10 CFR Part 100 guidelines. The PORV has two (2) possible failure modes, failure to open or, once open, failure to close. Following a DBA, alternate means, such as pressurizer safety valves and/or qualified high point vents, are available for RCS pressure relief if the PORV fails to open. Thus, the failure of the PORV to open would not jeopardize the capability to shutdown the reactor and maintain it in a safe shutdown condition. If the PORV fails to close following a signal to open, a motor operated block valve is available to isolate the flowpath through the PORV. In the unlikely event that both the block valve and the PORV fail to close, thus creating a non-isolable small break LOCA, analysis has shown the consequences to be acceptable under the licensing basis LOCA analysis. Again, the consequences of a failed open PORV would not jeopardize the capability to shutdown the reactor and . maintain it in a safe shutdown condition. For additional information concerning this area, please see Duke letter dated October 1, 1986 which was submitted in response to the NRC Safety System Functional Inspection Report dated August 1, 1986. Therefore, since the PORVs are not within the scope of 10 CFR 50.49 and since, as shown above, PORV failure would not jeopardize safety functions as outlined in 10 CFR 50.49 (b)(1), Environmental Qualification of the PORVs and associated control circuitry is not required. l
ENCLOSURE 1 MASTER FILE PT/0/A/201/04 Checked Control Copy Date/ Time Unit DUKE POWER COMPANY OCONEE NUCLEAR STATIO.N P.O.R.V. OPERABILITY TEST 1.0 Purpose To verify operability of the Power Operated Relief Valve RC-66 (P.O.R.V.) during heatup and prior to startup. 2.0 Referer.ces OFD-100A 3.0 Time Required 30 minutes 4.0 Prerequisite Test None 5.0 Test Equipment None 6.0 Limits and Precautions 6.1 Ensure that no personnel are in the Reactor Building S/G cavities while (1)(2)(3) RC-66 (P.O.R.V.) is being tested. 6.2 Do not conduct test at greater than 45 psig in RCS. 7.0 Test Method With steam bubble in Pressurizer and RCS pressure < 45 psig, l (1)(2)(3)RC-66 (P.O.R.V.) will be opened with (1)(2)(3)RC-4 (Power Relief Block) open and (1)(2)(3) RC-66 (P.O.R.V.) will be closed. Operator will verify opening of valve by observing a change in: Valve Outlet temperature, Pressurizer level, Quench Tank pressure, and P.O.R.V. Flow Monitor. I
PT/0/A/201/04 Page 2 of 4 Verify Date Date InitTime InitE7 Tune i 8.0 Data Required 9 8.1 Quench Tank pressure and temperature 8.2 P.O.R.V. outlet temperature 8.3 Pressurizer level 8.4 RCS pressure 9.0 Acceptance Criteria Test results are acceptable upon positive indication that (1)(2)(3)RC-66 (P.O.R.V.) opens and closes. 10.0 Required Unit Status 10.1 Plant in heatup mode with steam bubble in Pressurizer and RC pressure < 45 psig. 10.2 Quench Tank lined up for recircuhtion per OP/1,2, or 3/A/1104/17 (Quench Tank Operation). 11.0 Prerequisite System Conditions 11.1 (1)(2)(3)RC-66 (P.O.R.V.) outlet thermocouple ternperature available on computer. 11.2 Steam bubble f.n Pressurizer. 11.3 (1)(2)(3)RC-4 (Power Relief Block) operable. 12.0 Procedure 4 12.1 Verify that no personnel are in the S/G cavities in the Reactor Building. 12.2 Prior to cycling (1)(2)(3)RC-66 (P.O.R.V.), verify operability of (1)(2)(3)RC-4 (Power Relief Block) as follows: 12.2.1 Close (1)(2)(3)RC-4 (Power Relief Block). 12.2.2 Open (1)(2)(3)RC-4 (Power Relief Block). t. a i
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- g-PT/0/A/201/04 Page 3 of 4 64 t
{ .,q q' > / y. Verify ] ? >/ 'f i.L Y ~Date Date 1 3,b' Init 7Fime Init. / Time 3 tj ( 12.3 Pecord data required on Enclosure 13.1.(Information 1
- !!neet) prior to opening (1)(2)(3)RC-66 (P.O.R.V.).
J NOTE: . (1)(2)(3)RC-66 (P.O.R.V.) Key Operated Switch is (12.4) spring ret.prn to " AUTO" and must be held in "OPEN". '12.4 Open (1,)('2N3)RC-66 (P.O.R.V.) by using the Key Operated Switch in ICS Cabinet #13. / 12.5 Verify,(1)(2)(3)RC-66 (P.O.R.V.) open by: Quench TanN pressure increasing Quench Tank level increasing (1)d)(3)RC-66 (P.O.R.V- ) indicates open on Control Board. P.O.R.V. Flow Monitor indicates flow. Pressurizer. Relief Valve Flow Statalarr.t if 5 or a re lights are lit. C AUTION: If (1)(2)(3)RC-66 (P.O.R.V.) fails to close, (12 S) close (1)(2)(3)RC-4 (Power Relief Block) immediately. 12.6 As soon as there is positive indication that (1)(2)(3)RC-66 (P.O.n.V.) is open, close (1)(2)(3)RC-66 (P.O.R.V.) from ICS Cabinet #13 with Key Operated Switch by going to " AUTO". 12.7 Record data required on Enclosure 13.1 (Information Sheet) when (1)(2)(3)RC-66 (P.O.R.V.) is closed. 13.0 Enclosure 13.1 Information Sheet
Page 1 of 1 PT/0/A/201/04 ENCLOSURE 13,1 INFORMATION SHEET RCS Press. Quench Tank RC-66 PZR Thne < 45 PSIG-Press. Level Temp. Outlet Temp. Level l 4 I I I i t l )
ENCLOSURE 2 Pags No. 9 Report No. 47928-0 TABLE IV STEAM OPERABILITY AND LEAKAGE TEST, VALVE S/N BLO8903 DATE: Septerrber 12, 1985 i The valve inlet was pressurized to 45 + 5 psig and held for 30 minutes. The following tests were conducted. Leakage Test Inlet Pressure (psig) Pilot Leakage Main Leakage 45 + 5 Zero Zero Operability Test Run Inlet Pressure Response Time Steam Temperature Body Temperature No (psig) (msec) (*F) (*F) 1 45 + 5 155 303 256 2 45 T 5 170 300 256 3 45 _T 5 200 296 254 ' Leakage Test Inlet Pressure (psig) Pilot Leakage Main Leakage 45 + 5 Zero Zero Operability Test Run Inlet Pressure Response Time Steam Temperature Body Temperature No. (psig) (msec) ('F) (*F) 1 4 45 + 5 200 297 256 Leakage Test Inlet Pressure (psig) Pilot Leakage MainLegage 45 _+ 5 Zero Leakit,g 4 Pressure was increased to 500 + 10 psig and the valve temperatures stabilized. gakageTest Inlet Pressure (psig) Pilot Leakage Main Leakage 500 + 10 Zero Leaking WYLE LABORATOQlES ~ { Huntsville Facility -
v Page No. 11 Report No. 47928-0 ,.}- TABLE IV (Continued) STEAM OPERABILITY AND LEAKAGE TEST, VALVE S/N BLO8903 DATEi September 12, 1985 -) Additional Operability Test Inlet Pressure Inlet Pressure
Response
Steam Body ~ Run At Opening At Closing Time Temperature Temperature No. (psig) (psig) (msec) (*F) (*F) 75 638 534 9 2200 80 675 543 10 2650 '75 675 5 51 11 2650 ~ 75 671 553 12 2675 13-2675 2400 75 677 556 14 ~ 2675 2480 75 676 558 15 2675 2440 75 675 556 Leakage Test Inlet Pressure (psig) Pilot Leakage Main Leakage 2200 + 10 Leaking (Gross) Zero e 9 0 9
- s WYLE LABORATORIES Huntsvil!e Facility
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Page No.13 Report No. 47928-0 1 APPDEIX I CERTIFICATION TEST REPORT, S/N 8!.08903 l 3 ) 1 } l 1 ) d WYLE LABORATORIES Huntsville Facility c ~ ) O
( Page No.15 Report No. 47928-0 .1 WYLE 's"s?:s""""'s CERTIFICATION ) l LABORATORIES GROUP j Huntsville, Alabama 35807 TWX (810)726 2225, Phone (205)B37-4411 REPORT NO. 47928-1 } Mill Power Supply Company CUSTOMER P. O. NO. Ma4268-73 Ouke Power Company ) Oconee Nuclear Plant CONTRACT N/A Seneca, South Carolina NUMBER OF PAGES 6 DATE September 18, 1985 (
1.0 SPECIEN
Dresser Electromatic Relief Valve ) 2.0 PART NUMBER: 31533VX30 3.0 SERIAL NUMER: BLO8903 } i 4.0
SUMMARY
On September 14, 1985, valve serial number BLO8903 was tested for operc-bility and leakage with steam as the test medium. Testing was performed in accordance with Wyle Laboratories' Test Procedure 1011, Revision B, dated 1 April 3,1984, as modified by Duke Power letter from 0.H. Gabriel, dated October 24, 1984 The test data are tabulated below. 4.1 The inlet was pressurized to 45 psig and held for 1/2 hour. The following tests were conducted: LEAKAGE TEST In fat Pressure Pilot Main (psia) Leakace Leakaae 45 + 5 Zero Slfght
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- Page No.17 Report No. 47928-0 4.0 SUMMRY (Continued) 4.5 The inlet was pressurized to 2200 1 10 psig, the body temperature was . stabilized in accordance with the procedure, and the following tests were conducted: LEAKAGE TEST ] Inlet Pressure Pilot Nain (osio) Leakane Leakaa'e ] 2200 1 10 Zero Gross 4.6 OPERABILITY TEST Run Inlet Pressure Response Time Steam Temperature Body Temperature No. (psic) (msec) (*F) (*F) -{ ] 1 2200 + 10 105 650 563 2 2200 7 10 90 649 561 3 2200I'10 85 649 560 ~ LEAKAGE TEST Inlet Pressure Pilot Main (o<ia) Leakage Leakage 2200 1 10 Zero Slicht 5.0 VALVE REFURBISHMENT Valve S/N BLO8903 was disassembled and inspected. The main valve sprina was found collapsed about 1/4 of an inch. The pilot disc and pilot seat bushina was found with some call marks. The pilot seat bushing (seat end) also had some rouch areas (where the vertical wall and seat plane inter-sected). } The pilot disc was replaced with a new one from Duke Power. The new pilot disc was machined on the bottom end to allow for the spring to snap on. The seatir.a surf ace was also machined to the correct anale. The rough ~ areas of the pilot seat bushina were cleaned (machined) up. d The main sprina was replaced with a new one from Duke Power. (Prior to installation, the spring was compressed approximately 1/4 of an inch and heated to 500*F and held at this condition for one hour. It was allowed to cool down to ambient temperature and the tension released. The spring returned to its original length.) ~ The pilot and main disc were lapoed, the components were cleaned, and the valve was reassembled for testing. .o WYLE LABORATORIES r4u9tsydie Facibly /
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j ENCLOSURE 3 Page No. 5 Report No. 47140-0 7 TABLE II STEAM TEST DATA (S/N BLO8904), NOVEMBER 1, 1984 Run Inlet Pressure Body Temp. Delay No. (psig) (*F) (ms) Prior to Run 1 0 50.psig: Pilot Leakage = Zero Main Leakage = Leaking 1 50 252 180 190 ~- 2 50 250 3 50 230 4 50 After Run 4 @ 50 psig: Pilot Leakage =-Zero Main Leakage = leaking ~ 5 50 257 265 250 6 50 225 7 50 8 50 255 230 After Run 8 @ 50 psig: Pilot Leakage = Zero Main Leakage = Zero Prior to Run 9 0 500 psig: Pilot Leakage = Leaking Main Leakage = Zero 9 500 390 After Run 9 9 500 psig: Pilot Leakage a Zero Main Leakage = Zero 10 2200 535 95 11 2200 535 95 12 2200 534 90 After a 30-minute hold period, a leakaae check was performed at 2200 psig. Pilot Leakage = Zero Main Leakage = Zero x. .A -s e h l' ,e - I WYLE LABORATORIES l Huntsysile Facility b}}