ML20079H114
| ML20079H114 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/08/1982 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8212160044 | |
| Download: ML20079H114 (11) | |
Text
{{#Wiki_filter:-- DUKE POWER GOMPANY P.O. BOX 33180 CHAMLOTTE. N.O. 28242 HALII. TUCKER retzenoxz (704) 373-4831 var raramen? December 8, 1982 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414
Dear Mr. Denton:
On November 16-18, 1982, representatives from Duke Power Company, Westinghouse Electric Corporation, and the NRC Instrumentation and Control Systems Branch met in Bethesda, Maryland to discuss the open items which remained from previous meetings. Attached is a list of attendees and a meeting summary. Very truly yours, TJ bfA J Hal B. Tucker j ROS/php Attachment cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. P. K. Van Doorn NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq. Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 w$/ I Palmetto Alliance 2135 Devine Street Columbia, South Carolina 29205 8212160044 821200 PDR ADOCK 05000413 PDR A
Mr. Harold R. Denton, Director December 8, 1982 Page 2 cc: Mr. Henry A. Presler, Chairman Charlotte-Mecklenburg Environmental Coalition 943 Henley Place Charlotte, North Carolina 28207 Mr. Jesse L. Riley Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207 1 1 i i --,--,.,c. -.n,---.-
1 i l Meeting with Duke on 11/16/82 on ICSB Items i Name Organization f l K. N. Jabbour NRC/NRR/DL R. O. Sharpe Duke Power Co. D. W. Murdock Duke Power Co. T. A. Ford Duke Power Co. Dennis Serig NRC/NRR/DHFS/HFEB' Tom Dunning NRC/ICSB Fred Burrows NRC/ICSB P.M. Session (Additional Attendees) C. E. Rossi (Part-time) NRC/ICSB Nick Fioravante (Part-time) NRC/ASB Amarjit Singh (Part-time) NRC/ASB Faust Rosa (Part-time) NRC/IC3B I i a f i I i i i 4 4 i r -,.,-.y- ,3.- --,n .-w.-,- ,-.m., 7 .,._-~---e,, - -. -, - -.4-_ gn=,,, en, -.-- _e,-. -m-. 3 -,n
A t Meeting with Duke and Westinghouse on ICSB Items 11/17/82 i i Name Organization K. N. Jabbour NRC/DL/LBf4 T. A. Ford Duke /DE R. O. Sharpe Duke / Licensing Dennis Murdock Duke / Design Roy Owoc Westinghouse / Licensing Jack Mesmeringer Westinghouse /ICSL Ray Calvo Westinghouse / Control Systems Design and Analysis D. F. Dudek Westinghouse /F.S. Design Frank Orr NRC/RSB Tom Dunning NRC/ICSB Fred Burrows NRR/ICSB 2 f l i l 1 l 3 5 4 4 1 I j .. -,,. -.. ~. - - _ _. _.. - -.
wD 8 I NRC/ICSB - Duke Power Meeting Response Time Testing 11/18/82 i Name Organization R. O. Sharpe Duke Power Co. M. Virgilio NRC Jay Stackley Duke Power Co. Steve DeGange Duke Power Co. Marc Wigdor NRC 4 Tom Dunning NRC Fred Burrows NRC/ICSB H. C. Li NRC/ICSB K. Jabbour NRC i I l 4 J f k J i I i i 1 t e i
Duke - NRC/ICSB Meeting November 16-18, 1982 The following open items,were discussed: November 16, 1982 1. Lockout of Manual Control by Load Sequencer - Duke responded to the Staff's original concern; however, an additional concern was raised. The Staff agreed to write a letter detailing this concern. 2. Loss of RHR due to Single Instrument Bus Failure - Duke reiterated their previous position. The Staff was not satisfied, but agreed to pursue generically. 3. Control Rod Drive Shunt Trip Coil Testing - 1 Duke had previously indicated support for the Westinghouse position j and had requested that this item be pursued generically with Westinghouse. ] The Staff reclassified this item from a Technical Specification Item to an Open item. 4. Non-detectable Failure in Power Lockout Circuitry - Duke committed to test the circuitry in the closure direction at least once per refueling. This item is confirmatory. 5. NSW Pump Damage due to a Single Failure - i Following discussion, this item was clesed. 6. Auxiliary Feedwater System (AFW), TMI Item II.E.1.2, and JTl System Automatic Initiation and Flow Indication - These items were discussed with the Auxiliary Systems Branch and Reactor t Systems Branch reviewers in attendance. Duke proposed using the AFW block valves to control flow to the steam generators on a batch basis. Based on a valve stroke time of twenty seconds, the Staff found this approach acceptable. t 7. Remote Shutdown Instrumentation and Controls - i Duke agreed to provide a written response to the Staff's position. 8. Testability of Circt.itry for Transfer of NSW Suction from Lake Wylie to SNSWP - Duke explained the test program for the transfer of suction. It was agreed that this will be a Technical Specification Item.
9. Water Level Measurement Error - Duke committed to insulate the steam generator reference legs. This will be a Technical Specification Item. November 17, 1982 10. Installation and Testing of Automatic PORV Isolation System (TMI Item II.K.3.1) This item is under review by the Staff; no additional information is requited from Duke at this time. 11. Compliance with BTP RSB 5 This item is the subject of ongoing discussions between Duke and the Reactor Systems Branch. 12. UHI System - It was the Staff's position that no straightforward method was available for controlling the UHI System, specifically: a) The operator should have the capability to perform the safety functions, b) The operator has no direct indication of level in the UHI accumulator. c) The level sensing system is difficult to perform surveillance on. Westinghouse discussed the effect of the UHI System on a number of accidents analyzed in FSAR Chapter 15. Since the UHI System was designed as a passive system, no operator action is required. The Staff agreed to write a letter detailing any further concerns. November 18, 1982 Duke made a presentation on an alternate method of performing the Reactor Protection System Respons: Time Testing (handout attached). Since this method would not be used until the first refueling outage at the earliest, Duke agreed to submit the test procedure for ICSB review prior to the first refueling.
a DUKE POWER COMPANY CATAWD A NUCLEAR STATION CLOVER, S. C. 29710 . TELEPHONE: AREA 803 P. O. BOX 2 9 3 e3122e2 November 17, 1982 The following is submitted as an alternative to Reactor Protection System Response Time Testing as pursuant to IEEE 338-77, Periodic Test-ing of the Reactor Protection System. Baseline Data will be obtained on each protection system channel by in plant testing, manufacturer testing, or similar type testing to insure instrument channel response times are consistent with the assumptions made in the plant safety analysis., In addition to this, instrument channel trends will be obtained during the initial heatup and/or cooldown of the Reactor Coolant System in order to establish baseline instrument signatures. By periodically obtaining these channel signatures under similar initial conditions, a history of instrument channel performance can be obtained in order to determine if channel performance is consistent with the original baseline signatures and those of the redundant channels. Any instrument channel failures that would affect the response times during accident conditions, would also affect the performance during normal transient conditions and could be readily detected in the comparison of redundant channel performance. Deviation limits will be set based upon the initial channel signatures and corrective action taken based upon a detailed Engineering Evalua-tion. This may include channel calibration, Direct Response Time measurements, or other actions as deemed appropriate. Response time measurements will be made on protection system components whose response times are significant, that is, those compon-ents whose response times represent the majority of the total channel response time and those channels where signal analysis techniques can-not be performed. For those protectf.o pystem instrument channels that have intentional time delays, such as Reactor Coolant Pump under-voltage Reactor Trip, and overtemperature Delta T Reactor Trip, verifi-cations will be performed and documented in the respective channel I l calibration procedures, l To further support this approach to Reactor Protection Instrument l Response Time Testing, actual response times were obtained using a differential pressure transmitter with various diameters of impulse line piping to simulate the affect of clogging or plating on the interior of the pipe. A 50 foot impulse line was connected to a pressure sourse j monitored by a high frequency Validyne Test Transmitter with the trans-mitter under test connected at the end of the impulse line. A test signal equivalent to nominal Steam Cencrator level was applied then suddenly vented, simulating a low Steam Cencrator level Reactor Trig. 9
t e \\c s s m 3 \\* 7 November 17, 1982 - is - Page 2 a liigh speed recording equipmentywas utilized capable of resolving a '( tenth of a millisecond. The responic time was obtained for the sane input step, while the diameter of impulse line was varied. The res ilts - are as follows: 3/8 inch X 50 feet = 12 ms 1/4 inch X 50 feet = 19.5 ms., 3/16 inch X 50 feet = 21.ms As can be seen there is negligibic affect of i$ pulse line diameter on instrument response times. Tla actual impulse line's are 1/2 inch in diameter, larger than any of those tested. In a final' test an impulse line 600 feet long by 1/4 it.ch was connected and tile same pressure step applied. The measured response time was 156 milliseconds. Even in this grossly exagerater case transducer recponse was not signi-ficantly affected. Response time testing as,it pertains to the Reactor rotection System must be placed in perspective with regard to the-r.gnificance of a given components overali contribution to the total' channel response, ( time. The transducer and signal processing response tim +s are negligible compared to the total channel respon['e time. We therefore conclude that this approach is within the required scope of testice as outlined in IEEE 338-77, section 6.3.4., parapiaph 3, and that reliable protection channel performance can be mdintaisped.' 1 ky \\ t t l i N I I l l l ~' t k
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